ML20203C527
| ML20203C527 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 02/05/1999 |
| From: | Passwater A UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M98803, ULNRC-3957, NUDOCS 9902120110 | |
| Download: ML20203C527 (150) | |
Text
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union Electric ,
One Awren Plaza 1901 Chouteau Avenue PO Box 66149 Febmary 5,1$99 . st. iouis, MO 63166-6149 4 mannzt U.S. Nuclear Regulatory Commission ATTN: Document ControlDesk Mail Station PI-137 Washington, D. C. 20555-0001 ULNRC-3957 Gentlemen: TAC NO. M98803 DOCKET NUMBER 50-483 CALLAWAY PLANT UNION ELECTRIC COMPANY
[d- FOLLOW-UP ITEMS RELATED TO THE WA/11BM11 PROPOSED CONVERSION TO THE IMPROVED TECHNICAL SPECIFICATIONS SECTIONS 1.0,3.2. 3.3. 3.4. 3.6. 3.7. 3.8. 3.9. AND 5.0
References:
- 1. ULNRC-03578 dated May 15,1997
- 2. ULNRC-03889 dated August 27,1998
- 3. ULNRC-03877 dated August 4,1998
- 4. ULNRC-03927 dated November 25,1998
- 5. ULNRC-03900 dated September 24,1998
- 6. ULNRC-03853 dated June 26,1998
- 7. ULNRC-03905 dated October 21,1998
- 8. ULNRC-03937 dated December !1,1998
- 9. ULNRC-03908 dated October 21,1998
- 10. ULNRC-03926 dated November 23,1998
- 11. ULNRC-03946 dated December 22,1998 Union Electric Company requested an amendment to the Callaway Facility Operating License (NPF-30) by incorporating changes to the Technical Specifications (TS) as provided in Reference 1. The NRC staff requested additional information regarding Section 1.0, "Use and Application," Section 2.0,
" Safety Limits (SLs)," and Section 3.0, " Limiting Condition for Operation (LCO)
Applicability and Surveillance Requirement (SR) Applicability," which was provided in Reference 2. The NRC staff requested additional information regarding Section 3.1, " Reactivity Control Systems," Section 3.2, " Power Distribution Limits," Section 3.5, " Emergency Core Cooling Systems," Section 3.9, " Refueling Operations," and 4.0, " Design Features," which was provided in Reference 3. The NRC staff requested additional information regarding Section 3.3, "Instmmentation," which was provided in Reference 4. The NRC staff requested additional information regarding Section 3.4, " Reactor Coolant System,"
and Section 5.0, " Administrative Controls," which was provided in Reference 5.
f,N e The NRC staffrequested additional information regarding Section 3.6,
" Containment Systems," which was provided in Reference 6. The NRC staff '
I requested additionalinformation regarding Section 3.7," Plant Systems," whic was provided in Reference 7. In addition, the NRC staffrequested additional e information regarding Section 3.8, " Electrical Power Systems," which was provided in Reference 8.
9902120110 990205 "
DR ADOCK 0500 3
- U.S. Nuclear Regulatory Commission February 5,1999 ;
Page 2 l
The Attachments to this letter provide (1) additional information or supporting documentation not provided in the original RAI responses, (2) answers to follow-up questions, and (3) additional changes identified by the licensee for ITS Sections 1.0,3.2,3.3,3.6, 3.7,3.8,3.9, and 5.0. Please note that Item Number Q 5.2-1, TSTF-258 is applicable to Callaway Plant, but because of incomplete internal review, we will provide our response to this item in the next l follow-up letter. Reference 9 provided the first follow-up letter, affecting ITS i Sections 3.1,3.2,3.4,3.5, and 5.0. Reference 10 provided the second follow-up i letter, affecting ITS Sections 1.0, 3.4, 3.5, 3.6, 3.7, and 3.9. Reference 11 provided the third follow-up letter, affecting ITS Sections 1.0, 3.1, 3.2, 3.3, 3.4, 3.7,3.9, and 5.0.
i This letter and its Attachments are not a supplement to Reference 1 and have not been reviewed by the Onsite Review Committee or Nuclear Safety Review Board. A supplement to Reference I will be provided at a later date.
Ifyou have any questions concerning this response, please contact us.
Sincerely, Ya an C. Passwater c
Manager, Corporate Nuclear Services ACP/pir i Attachments l l
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= . .-_ _ . _ _. . . _ _ . _ _ . - . ._ .. _ . ______ _. _ _ _ _ _
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j STATE OF MISSOURI )
i 1
Alan C. Passwater, of lawful age, being first duly sworn-upon oath says that he is Manager, Corporate Nuclear Services for Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
By OM Alan C. Passwater ;
Manager, Corporate Nuclear Services SUBSCRIBED and sworn to before me this Md.- day of 4Q4Wm7u , 1999.
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5Y PATRICIA L REYNOLDS Nowly PuBUC-8TAEOF MMBOWB ST. Lout 8 COUNTY MV - DRM08.31,MS J
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cc: M. H. Fletcher Pr. fessional Nuclear' Consulting, Inc.
19 41 Raines Drive Derwood, MD. 20855-2432 Regional Administrator b/A U.S.-Nuclear' Regulatory Commission Region IV 611 Ryan Plaza Drive JSuite 400 Arlington, TX 76011-8064
. Senior Resident Inspector W /'
Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Mel Gray (2) W/A Office of Nuclear Reactor Regulation
~U.S. Nuclear Regulatory Commission l' White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 Ron Kucera Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Denny Buschbaum . ts /w TU Electric P.O. Box 1002-Glen Rose, TX 76043 Pat Nugent W h-Pacific-Gas & Electric
-Regulatory Services P.O. Box 56 Avila Beach, CA 93424 l
I l
i i
i The item numbers are formatted as follows: [ Source][lTS Section}-[nnn)
Source = Q - NRC Question NR - NRC Follow-up Question CA- AmerenUE DC-PG&E WC - WCNOC '
CP - TU Electric TR - Traveler ITS Item Number - Applicability Enclosed !
1.0 Q 1.1-1, Q 1.1-2, TSTF-205 CA, CP, DC, WC YES l 3.1 CP-3.1-004 (NEW) CP NA !
3.1 CP-3.1-005 (NEW) CP NA )
3.1 CP-3.1-ED1 (NEW) CP NA 3.2 0 3.2-6, TSTF-241 CA, CP, DC, WC YES 3.3 0 3-LS-GEN, TSTF-135 CP NA l 3.3 O 2-13 CP NA l 3.3 Q 2-32 CP NA l 3.3 O 8-04 CA, CP, DC, WC YES 3.3 O 3.3d GEN CP NA 3.3 0 3.3;i _ _ CA, CP, DC, WC YES J.3_ G 3.3-24 CP,DC NA 3.3 Q 3.3-55 CA, CP, DC, WC YES
_3.3 0 3.3-63 CA, CP, DC, WC YES 3.3 0 3.3-73 CP NA 3.3 Q 3.3-78 CP NA 3.3 0 3.3-96 CA, CP, DC, WC YES 3.3 Q 3.3-97 CA, CP, DC, WC YES 3.3 CP-3.3-ED1 (NEW) CP NA 3.3 CP-3.3-ED2 (NEW) CP NA 3.3 CP-3.3-ED3 (NEW) ,CP NA 3.3 CP-3.3-ED4 (NEW) CP NA 3.3 CP-3.3-ED5 (NEW) CP NA _ j 3.3 CP-3.3-ED6 (NEW) CP NA ;
3.3 CP-3.3-ED7 (NEW) CP __
NA :
3.3 CP-3.3-002 CP NA )
3.3 CP-3.3-009 CP NA 3.3 CP-3.3-010 CP NA 3.3 CP-3.3-011 CP NA i 3.3 CP-3.3-012 (NEW), TSTF-169 CA, CP, DC, WC YES 3.3 CP-3.3-013 (NEW) CP NA i i
3.3 CP-3.3-014 (NEW) CP NA 3.3 CP-3.3-015 (NEW) CP NA >
3.3 CP-3.3-016 (NEW) CP NA 3.3 CP-3.3-017 (NEW) CP NA 3.4 Q 3.4.11-4, TSTF-309 CA, CP, DC, WC YES :
3.4 O 3.4.12-2, TSTF-285 CA, DC, WC YES l
3.4 CP-3.4-005 (NEW) CP NA I i
3.5 CP-3.5-005 (NEW) CP NA 3.6 O 3.6.3-23 CP,DC NA 3.6 O 3.6.3-54 CA, WC YES 3.6 CA-3.6-ED1 Mv/) CA YES 3.6 CF 3.6-ED (NEW) CP NA 3.7 0 3.7.3-2 CP NA 3.7 O 3.7.5-4 WC NA 3.7 O 3.7.16-3 _
CA, WC YES 3.7 CP-3.7-ED CP NA 13.8 0 3.8.1-20 CP CP' NA 3.8 0 3.8.2-02 CP CP NA 3.8 Q 3.8.4-19 CP NA 3.8 Q 3.8.10-01 CA CA, CP, WC YES Q 3.8.10-01 CP Q 3.8.10-01 WC 3.8 O 3.8.8-02.p CP NA 3.8 CP-3.8-ED CP NA 3.8 CP-3.8-005 (NEW) CP NA 3.8 CP-3.8-006 (NEW) CP NA 3.8 CP-3.8-007 (NEW) CP NA 3.9 Q 3.9-1a, TSTF-272 CA, CP, DC, WC YES
~
l 5.0 Q 5.2-1, TSTF-258 CA, CP, DC, WC NA*
I 5.0 0 5.5-2, TSTF-237 CA, CP, DC YES I
. 5.0 _
I
- Information will be provided in the fifth follow-up letter.
I ADDITIONAL INFORMATION COVER SHEET I
ADDITIONAL INFORMATION NO: O 1.1-1 APPLICAB;LITY: CA, CP, DC, WC REQUEST: CTS 1.3, Analog Channel Operations Test CTS 1.3, Channel Operational Test (Diablo Canyon)
CTS 1.7, Channel Functional Test (Diablo Canyon] 1 CTS 1.35, Trip Actuating Device OperationalTest [ Wolf Creek) l l
CTS 1.36, Trip Actuating Device Operational Test (Callaway] I CTS 1.37. Trip Actuating Device Operational Test [ Comanche Peak) {
CTS 1,38, Trip Actuating Device Operational Test [Diablo Canyon]
DOC 1-30-A l ITS 1.1, Channel Operational Test (COT)
ITS 1.1, Channel Functional Test (CFT) [Diablo Canyon]
ITS 1.1, Trip Actuating Device Operational Test (TADOT)
JFD 1.1-9 These are changes to both the CTS and the STS and are considered generic. l Therefore, they are beyond the scope of the conversion review. The DOC states that I these changes are consistent with TSTF-39, Rev.1. Also, Diablo Canyon's ITS markup appears to be in error as shown by " Channel Operational" versus " Channel Operational ,
Test (COT)." !
Comment: If NRC has not approved TSTF-39 by the time the draft safety evaluation is prepared, then these changes should be withdrawn from the conversion submittal at that time. These changes will not be reviewed on a plant-specific basis. In addition, correct the Diablo Canyon ITS markup for COT.
FLOG RESPONSE (original): TSTF-39 Rev.1 has been withdrawn by the TSTF. However, changes addressed in TSTF-39 Rev.1 have been subsumed by TSTF-205. Revision 1 of TSTF-205 is currently undergoing final review by the TSTF members. After final wording changes for the COT definition as well as Section 3.3 Bases changes to establish the requirements for relay contact surveillance testing (issue originally raised at Peach Bottom) are agreed upon, Revision 1 will be submitted for NRC review. The attached pages represent the definition changes to be included in Revision 1. The Section 3.3 Bases changes will be addressed under that Section's review.
FLOG RESPONSE (supplement): The FLOG proposes to incorporate Revision 3 of TSTF-205 which has been approved by NRC. The only changes required involve the definitions of l MASTER RELAY TEST and SLAVE RELAY TEST. ITS Section 3.3 Bases changes to relax the requirements for relay contact surveillance testing, which per the traveler may be l performed during COTS and TADOTs by the verification of the change of state of a single l contact of the relay, have not been incorporated.
1 ATTACHED PAGES:
Attachment 4, CTS 1.0 - ITS 1.0 Enclosure 2, pages 1-3 and 1-5 Enclosure SA, Traveler Status Sheet Enclosure SA, pages 1.1-4 and 1.1-6 l
l-
~0EFINITIONS '
a ep.\
c.g:, E - AVERAGE DIS!NTEGRATION ENERGY li 1. :'- T shall be the average (weighted in' proportion to the concentration of I each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gama energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of tne total noniodine activity in the coolant.
(ESF)
V ENGINEERED SAFETY FEATURES RESPONSE TIME
': The C;;;;;;"n"0 :Ar"TY T:A
- SF RESPONSE TIME shall be that time l interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its /-d/-d safety function (i.e., the valves travel to their required positions, pumo discharge pressures reach their required values, etc.). Times shall include l
diesel generator starting and sequence loading delays where applicable.g
~
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- DE F:"'"O LEAXAGE p 3*,4 ' ' ' ' '
,4 ,,KAK,A .. G.d..ci,*//.Le :
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LEAXAGE_
/-// j A
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/,-a,- L; .i;;; (aerp. 00t T"0LL"" L:^1"CE) ' .t; ;1;;-d y;t; .. ,"such as #,fM g ca th;t r; M ptured rs- and conducted to ca// sed ~ea ry.rdew t E4XACK on pump sealtor salve packing f
i;;i:a su '
.'2 -l>:- Ltd;;;"into the containment atmosphere from sources that are both
- specifically. located and known either not to interfere with the operation of Leakage Detection Systems or not to be "":!!"".E ?D"uDAov LEAXAGE; or
($ct) 5EAXACK fr-*//mee m a d * )
Reactor Coolant System l::k.,e through a steam generatorV J'. .crr- a to the
, Secondary C;;;.ui. System; (cg
- 1~N J6M7~ /,/- s~ g&& l MASTER RELAY TEST I con.rir+o-f . e f.-pj-A
- . Q A MASTER RELAY TEST shallD : " energizY: D . ;f r =
- e and /-ykA ;
verifi;;tien ;f OPERABILITY of eachMlay. The TER RELAY TEST shall include l acofinuit check of each associatedgslave rela ."TXs /hArrE4 #E(.A TEr/~ d /,/-/
Q "E"et. L cr m : u:L::
V egired Lefsehrmed meanc any .ra na.c o f .r huo lyfmp av hk/4ep.e. <gaa, Mig _,2_ ;
- . 7 MD :?.( ' 07 TH: "U"L:: she::
1 W - ?!y :::::ia::d ::1" :h: ?? :t. ~'f: ::' incl. deep;,per;;n;wh;;r;not;;;;p:
y d;; act in-h.d :.-7 y::
cf tn; li;;nsee, it; ;;ntre;ter; ;r .;nder Ah ; c n kd:d fr;- thi: ::teg:ry
- p;r;;n: ip.;nt ;r :: =he d;1inri::.
Thi ce ;;ery dee; in;;.de persen; n; .;a n; cat;r th; :it; t; ,;rsin cg;rtien; ;f'th; ;it; for r::re:-
- %, ::', , :::.;;: \. n:: , :. ::t:r ;.r ;::::
p,:::::':::1 "'.!t :t: :':~' .
-.' all ne r niniin hke che**l 13 tut ehu- Ahe o.cMnnel n . *N 0ADtA8tL1~rf CALLAWAY - UNIT 1 -
. - . . . ..._m_.
I DEFfNfTf0NS .
QUADRANT POWER TILT RATIO [&f r4)
C..m /-/P-A v; .. " " 25 0'.'".0"".."T "C'.JER TILT RATWhall
.pp-r gbe the ratio of the maximum upper excore l
. detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of .the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, which-ever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATEDTHERMALPOWER(#1/)
A11' 1.25 ""TCO TU;LA. T^WE"J'shall be a total core heat transfer rate to the
. /-61-lA reactor coolant of 3565 MWt. l
. (A1T) \
l REACTOR TRIP SYSTEMVRESPONSE TIME RTT (~ /-p7-A !
-i.G The R:A" OR TR:" :YST "" RESPONSE T1 ME shall be the time interval from ,
when the monitored parameter exceeds its1 Trip Setpoint at the channel sensor i until loss of stationary gripper coil voltage. .ZWrEg7-/./-G f) /, /4 1
e...enn orin , e eu. r.u. .-
4" 1.:S A R "0"Ta"L" :" "T :h:11 5: :n3 :f th::: ::Miti^": tr-"4-d l
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.chichreer i: 1:::. A"L"" :nd A"L * "" :r: defk:d i, !;::i'i::ti::: 2.2.'
- nd 4.2.2.2, r;; p ;tiv;1y. "A"00 ;;j 5 :nter:d :t the de:reti:r :# th: -
1i en:::.
SHUTDOWN MARGIN [fbM) m <- cank l (gccAr) j- y 1 . 2 0 -:"';- ~~^'. " F,'" ; Wshallbetheinstantaneous(amountofreactivitybywhich l
the reactor is suberitical or would be suberitiqa1 from i'J present condition assuming all 'ui' ':n;;th red clustemassemblies":h td:,7 :nd : ntr:1-} are fully inserted except for the singl :d :12:t:r :::x.tij-of highest reactivity worth which is assumed to be fully thdrawn. 2 AltEM / /-//
8CCA
- e. r. . r. o n, ,u n n.ov
/-2+-A line bey:nd hich the 1:nd i' 1.01 Th: !!T 00"CARY sheli : the V.Y
- th r l---
Cr.:d, n:r '. :::d, n:r :th:mi:: ::nt ell:d by th: en et.
all s fg,j& a r*(f id SLAVE RELAY TEST r' '
co neir + d in oI-A 4-M- A SLAVE RELAY TEST shall% the energiz:t.:r ;f- -
V r a and /-/-30-A veri S:ti: :P OPERABILITY of each relay. The SLAVE RELAY TEST shall includ /./-/
a; co tinuit check =" : ~4 "# r o s s%ci dg tes le actuation device yT3 -Se E h* Y Al n o y 1-5 Amendment No. U , $8, 2g , 50 CALL A - UNIT 1 7he MM f AGtA)'}m' TE71~M*;r 4efee$arme) y l nem'af **y teeier oA reguen e veeloff y oo.-fy / r/ejr,
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. I INDUSTRY TRAVELERS APPLICABLE TO SECTION 1.0
,g. TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-19, Rev.T1~ @ Incorporated @ /, /-/2, @ NRC approved, I umewse 72-/M
! ~
2ni Aar.3 i,i 1.1-9 i
!- S TSTF w A Incorporated Ngc approve /. Q /,/-/ l
& W G & /,/-2 m re GD ,
2 /,/-9 I TSTF 111, Incorporated 1.1 5 l l
Rev. Q S' f,j I
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Incorporated 1.1 6 MACS /MPel, $/, /-d'
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Cncer,wded chufer- $3.4.1-f, NRC cM, I
I i
i MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97
- . , . ~.
- .- - . _ . . l Definitions 1.1
, .q.y J
(g. 1.1 Definitions LEAKAGE 2. LEAKAGE into the containment atmosphere from i (continued) sources that are both specifically located l and known either not to interfere with the l operation of leakage detection systems or not I
\
to be pressure boundary LEAKAGE: or i
- 3. Reactor Coolant System (RCS) LEAKAGE through ;
a steam generator (SG) to the Secondary System:
- b. Unidentified LEAKAGE ;
All LEAKAGE (except RCP seal water injection er 1.1 10 I leakoff) that is not identified LEAKAGE: ,
LEAKAGE (except SG LEAKAGE) through a nonisolable !
fault in an RCS component body. nina wall . or vaua wail. DIl N Mi*?s in Se ch****l .
L refuTo _ e c W ne/
MASTER RELAY TEST A MA3TER RELAY TEST shall consist of energizind(ggD I /./-9 l
(([mp(erceGb and verifying the OPERABILITY of each relay. M /.C 3:3' '
~
The MASTER RELAY TEST shall include a continuity check of Ol*/-/ j
_ each associated 4
reguiredslave r)elay. ,Y,7,.,g g,. j H00E A H0DE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.11 with fuel in the ;
reactor vessel. '
OPERAB OPERABILITY A system, subsystem, train. component, or device shall be OPERABLE or have OPERABILITY when it is capable of gg performing its specified safety function (s) and when all
[ / [7 'necessary attendant instrumentation, controls, normal or
'II/ *" emergency electrical power, cooling and seal water, e dag.re,-ref ,8eefu=dA/ lubrication, and other auxiliary equipment that are e ven-/egry. ,, f,Q,,/ f./, equired for the system, subsystem, train, component, or
- n 't / ,, jj , ,4 / device to perform its specified safety function (s) are ,
, 7 i "f" Ej also capable of performing their related support 6,7 ,
, 7 function (s).
(continued)
HARK-UP OF WOG STS REV 1 (NUREG 1431) 1.1 4 5/15/97
_ _ _ ~
- - - , a-.. - - a l
Definitions 1.1 p
'y 1.1 Definitions SHilTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity oy which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All rod cluster control assemblies (RCCAs) are fully I inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b. In H0 DES 1 and 2, the fuel and moderator temperatures are changed to the ret.inci zccc pc.;;r 3,p3' desi;in itsel STAVE RELAY TNT < - -
r// t/ eve rek7 r fr#2 c/wnne/
re zrre/ Ar eb/ d,t/
A SLAVE RELAY TEST shall consistf,'f_f gnd verifying the OPERABILITY of eacgslave lay.
of ene l TliE SLAVE RELAY TEST shall include, sr 51 i .
" continuity check of associated testable ac uation evices.
& 4 retwred reguired)
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsys%s, channels, or other designated j components during the laterval specified by the 7X, J2AvcgetA77 Err %.y Surveillance Frequency, so that all systems, subsystems, le foe M e//7 Aeav sFn7 l nchannels, or other designated components are tested during Surveillance Frequency intervals, where n is the total
/
feg,f,.f.reguen#.M/
a ve,-/, ,r 4 s, number of systems, subsystems, channels, or other
/; /',, 7, t p// - e designated components in the associated function.
> )r a _ -- d 1, _
RHAL POWtx THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. i e hke TRIP ACTUATING DEVICE A TADOT shall consist ofLoperating the trip actuating J OPERATIONAL TEST device and verifying thev0PERABILITY(cf rcquired alarm, 1.1 1 (TADOT) ....~ ~. misplay and trip functicit. .. ,. m a 1.1-9 l n.f a ll evicer in +Le _ are,o we- a n w w = -
> - - ~-
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The TADOT shall include adjustment, as necessary, of the l trip actuating device so that it actuates at the requiredy setpoint within the -requiref 4accuracy.4 ntCCJJarf MARK UP OF WOG STS REV 1 (NUREG 1431) 1.1 6 5/15/97
ADn!TIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 1.1-2 APPLICABILITY: CA, CP, DC, WC
~ REQUEST: CTS 1.3, Analog Channel Operations Test CTS 1.5, Channel Calibration i CTS 1.35, Trip Actuating Device OperationalTest [ Wolf Creek) I CTS 1.36, Trip Actuating Device Operational Test [Callaway) j CTS 1.37,-Trip Actuating Device Operatiend Test [ Comanche Peak) .j CTS 1.38, Trip Actuating Device Operational Test [Diablo Canyon) - 1 DOC 1-32 A :
ITS 1.1, Channel Calibration I ITS 1.1, Channel Operational Test (COT) )
ITS 1.1, Trip Actuating Device Operational Test (TADOT) l JFD 1.1-1 {
These are changes to both the CTS and the STS and are considered generic changes. )
Therefore, they are beyond the scope of the conversion review. The DOC states that ;
these changes are consistent with TSTF-64. I J
Comment: If NRC has not approved TSTF-64 by the time the draft safety evaluation is l prepared, then these changes should be withdrawn from the conversion subr9tal at that time. These changes will not be reviewed on a plant-specific basis i
- FLOG RESPONSE (original): TSTF-64 has been withdrawn by the TSTG. However. mangas addressed in TSTF-64 have been subsumed by TSTF-205. Revisbn 1 of TSTF-205 is i currently undergoing final review by the TSTF members. After 'inal wording changas for the a COT definition as well as Section 3.3 Bases changes to establisq the requiremerus for relay ' !
. contact surveillance testing (issue originally raised at Peach Boticm) are agreed upon,-
. Revision 1 will be submitted for NRC review. The attached pages represent the definition ;
- changes to be included in Revision 1. The Section 3.3 Bases changes will be addressed under i that Section's review. l
}
i F' CO RESPONSE (supplement): The FLOG proposes to incorporate Revision 3 of TSTF-205 !
which has been approved by NRC. The only changes required involve the definitions of l MASTER RELAY TEST and SLAVE RELAY TFiST. ITS Section 3.3 Bases changes to ;
relax the requirements for relay contact surve.alance testing, which per the traveler may be l performed during COTS and TADOTs by the verdcation of the change of state of a single j contact of the relay, have not been incorporated. See also the response to Comment Number !
Q 1.1-1.
l A1TACHED PAGES:
t None f
f E
)
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.2-0 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.2.4 Quadrant Power filt Ratio CTS 3/4.2.4 Quadrant Power Tilt Ratio (All FLOG Plants)
DOC 04-01-A JFD 3.2-05 ITS Required Action A.5 Comment: The ITS proposes to change the STS wording for Required Action A.5 from
" Calibrate excore detectors to show zero QPTR," to " Normalize excore detectors to eliminate tilt," based upon WOG-95 (and rejected TSTF-25). A preferred wording would be that proposed in the Comanche Peak CTS mark-up, " Calibrate excore detectors to .
show zero Quadrant Power Tilt." What is status of WOG-957 FLOG RESPONSE:
Traveler WOG-95 was transmitted to the NRC in February 1998 as TSTF-241. The l FLOG is incorporating TSTF-241 including the latest revisions discussed at the ,
June 1998 WOG MERITS Mini-Group meeting. These revisions corrected errors made ,
during the development of TSTF-241.
I Additionally, Wolf Creek submitted a License Amendment Request to CTS 3/4.2.4, Quadrant Power Tilt Ratio, on February 4,1998 which was approved on April 27,1998 in Amendment No.116. This amendment incorporated the changes proposed in .
TSTF-241. l The FLOG believes that it is appropriate to incorporate the proposed TSTF-241 changes based on the NRC approval of the Wolf Creek amendment request.
FLOG RESPONSE: (supplement)
The NRC and the Technical Specification Task Force have agreed upon the acceptable changes to be included in this traveler. Most of these changes were incorporated into the FLOG submittals in the initial RAI response as noted above. The final set of needed modifications have been incorporated as noted in the attached pages.
FLOG RESPONSE: (supplement)
This response is being supplemented to correct editorial errors made in the inserted text for Bases 3.2.1, Actions A.2 and A.3. Additionally, TSTF-241, Rev. 4 was recently issued and is believed to be the Revision to be approved by the NRC. Revision 4 had been accurately reflected in previous FLOG responses. The Traveler Status page is only being updated to reflect the issuance of Revision 4.
l l AT1 ACHED PAGES: t Att. No. 8 CTS 3/4.2 -ITS 3.2 Encl.5A Traveler Status page Encl. 5B B 3.2-5 i
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- 4a na Zndustry Travelers Applicable to Section 3.2 l A,yL.Ala b cAs c only (cper.).
.n
([;Q TRAVELER # stall)S DIFFERENCE #
/' COMMENTS TSTF-24 Not Incorporated NA Not NRC approved as of gev /
ytraveler cut off date.
TSTF-95 Incorporated 3.2 06 Approved by NRC.
TSTF-97 Incorporated 3.2-07 Approved by NRC.
TSTF 98. Incorporated 3.2 03 Rev. + 2 -TE 7.Q-042 TSTF 99 Incorporated 3.2 08 Approved by NRC. I TSTF 109 Incorporated 3.2 15 Approved by NRC.
TSTF 110 Incorporated 3.2 10 Rev. + 2 Afgewe/ 37 N/Ld, -n6-7.2-4df TSTF-112, M Incorporated -H#- O t "nc =n ~ ed r c' Rev. 1 7.2-2a[> 77jele7 ft g<, ago W.2-e3 TSTF 136 In'corporated NA A rp =- J g gAc. -rg-F.:2-@jt-idir 164 Incorporated 3.2 11 Applicable to CAOC only (CPSES).
iCC 23. Rev. ^ . nco ated 3.2 05 ' 'il' dr.ccrper:tc g
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MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97
_ _ _ _~
Fe (Z) (F aMethodology)
B 3.2.1&
.A
- f. 4
%$# BASES ACTIONS L.1 (continued) orderly manner and without allowing the plant to remain in ag -
y A.] unacceptable condition for an extended period of time. /d?3.r.-6) r - **K.
) c g 3.2.-o02--.
a -
A reduction of the Power Range Neutron Flux-High trip setpoints by 21% for each 1% by which Fj(Z) exceeds its 1 Nit, b a conservative action for protection against the consequences of k severe transients with unanalyzed power distributions. The Completion Time of 8 Z2 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding' prompt reduction in THERMAL POWER in accordance with s Required Action A.1 9 7 g 3,z_. [
L.3.
Reduction in the Overpower AT trip setpoints by 11% for each it gl by which Fj(Z) exceeds its limit, is a conservative action for
, protection against the consequences of severe transients with
'~~ unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i is sufficient considering the small likelihood of a severe .
transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1 9 w A.3 _
Q *i . 2
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Verification that Fj(Z) has been restored to within its limit, by l performing SR 3.2.1.1 prior to increasing ViERMAL POWER above the limit imposed by Required Action A.1. ensures that core !
conditions during operation at higher power levels are consistent {
with safety analyses assumptions. InheFsnt"'in3 hts',TctTonlis !
LdsstJffat,i on'.Tol .theliUssWfat,he~_ cytGiofjl i mitZr.onditjyfian_d
. l the]co rre_cti onlof-thicaus eBtolthe~ bgenfine,ce ss a ryftoC all o'w ;
saff operation?at'the higher power:le~ vel'.J"The-61Towableypdwer !
l evel 11 s! determined".,bEeyaluat_ing (Fj(Rifor3he,]11 gh'e,r2 power !
),ep,e_1..T., 3R3'l2'El' must3bsishtisfiedVrMr,it6Tindrea.si5g?pcwir above :the' higher 7al]owiblCpower ]evel 'or restoration of any -
reducediReactoC_ Trip System"Setpojnts,;
1 (continued)
MARK UP OF NUREG.1431 BASES B 3.2-5 5/15/97
-. wm,+ we.+. . -%. i l, O 3.2-6 !
I
.c 5 INSERT A.1 r3 ;3 M5N . The maximum allowable powerlevelinitially determined by Required Action A.1 may be affected by subsequent determinations of F8(Z) and would require power reductions within 15 minutes of I the F8(Z) determination, if necessary to comply with the decreased maximum allowable power !
level. Decreases in F8(Z) would allow increasing the maximum allowable power level and )
increasing power up to this revised limit. '
INSERT A.2 - t The maximum all e Power Range Neutron Flux - High trip, setpoints initially determined by Required Action. riay be affected by subsequent determinations of F8(Z) and would require I i Power Range Ne ron Flux - High trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the F8(Z) I' l determination, if necessary to comply with the decreased maximum allowable Power Range l Neutron Flux - High trip setpoints. Decreases in F8(Z) would allow increasing the maximum allowable Po%er Range Neutron Flux - High trip setpoints.
I INSERT A.3. 3
\
The maximum allowable Overpower AT trip setpoints initially determined by Required Action may be affected by subsequent determinations of F8(Z) and would require Overpower AT trip j setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the F8(Z) determination, if necessary to comply with the !
decreased maximum allowable Overpower AT trip setpoints. Decreases in F8(Z) would allow !
- increasing the maximum Overpower AT trip setpoints. ~
]
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l ADDITIONAL INFORMATION COVER SHEET l
l ADDITIONAL INFORMATION NO: Q 8-04 APPLICABILITY: CA, CP, DC, WC REQUEST: Consistent with NUREG-1431, Rev.1 (ITS 3.3.3 Required Actions C.1, E.1, and G.1), this change deletes the requirement to initiate an alternate means of monitoring within l 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when two channels of containment radiation level [or RVLIS) are inoperable as specified in current TS [3.3.3.6 ACTION c].'
Comment: What is the basis for deleting the requirement to initiate the alternate method of monitoring in the case of containment high range radiation monitoring? The NSHC indicates that this automatically happens in the case of RVLIS, but implies that i PASS monitoring of the containment is not initiated under condition G. i FLOG RESPONSE (original): NSHC LS-17 has been revised as per the attached.
l FLOG RESPONSE (supplement): NSHO LS-17 has been further revised to provide examples of instrumentation available to detect inadequate core cooling.
ATTACHED PAGES: 1 Attachment 9, CTS 3/4.3 - ITS 3.3 {
Enclosure 4, page 49 .
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.2 IV. SPECIFIC NO SIGNIFfCANT HAZARDS CONSIDERATIONS Ed - NSHC LSJ17 10 CFR 50.92 EVALUATION l
^
FOR- 'I TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREENTS WITli!N'THE TECHNICAL SPECIFICATIONS This change deletes the requirement to initiate the preplanned alternate method of -l monitoring containment radiation [or reactor vessel water level] within:72 hours ;
~ when two channels.are' inoperable. This makes ACTION [c] of current TS LCO [3.3.3.6] j the same as ACTION _ [b). except that a plant shutdown is not needed if the 7. day A0T - ~j
.is not met for these functions with preplanned alternates. Containment radiation l indication is used to assess whether the fuel cladding and reactor coolant pressure j
, boundaries have been breached such that a significant portion of the core activity ')
inventory is available for. release to the environs. The preplanned alternate- for !
this' variable uses the PASS system which is administrative 1y controlled outside the
. TS. [The preplanned alternates for RVLIS include a combination of the core exit i
t les. RCS wide range hot and cold leg temperature, wide range RCS pressure. .
lpri ssurizer level, and RCS subcooling monitor indications to verify adequate core !
211ng.] These variables are already continuously monitored under the LCO and do j t ire an action directing that they be initiated. l SEtrLE-11 & P-bH.E-l"7 l The proposed TS change has been evaluated and it has been determined that it !
involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in.10 CFR 50.92(c)' as quoted below- I l
L"The Commission may make a final determination. pursuant to the procedures in 50.91, that a proposed amendment to an' operating license for a facility .,
licensed.under 50.21(b) or 50.22 or for a testing facility involves no significant hazards consideration, 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 the possibility of a new or different kind of accident from any j accident previously evaluated; or j
- 3. Involve a significant reduction in a margin of safety."
The following evaluation is provided for the three categories of the significant j hazards consideration standards:
- 1. Does the change involve a significant . increase in the probability or ;
+
consequences of an accident previously evaluated?
Overall protection system performance will remain within the bounds of the
. previously performed accident analyses since no hardware changes are proposed.
l NO SIGNIFICANT HAZAPDS CONSIDERATION 49 5/15/97
,s-ma . y n.,a.-en-; -,rnm---
p~
l INSERT LS-17 Q 8-04 !
i t
Implementation of containment air sampling via the PASS will be initiated as required by ITS 3.3.3 Condition G.' There is no safety benefit to activating the PASS earlier than required by Condition G considering the other highly reliable indications available to detect inadequate core cooling, such as Class 1E core exit thermocouples, RCS wide range hot and cold leg ,
temperature, wide range RCS pressure, pressurizer level, and RCS subcooling monitor !'
indications.
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l ADDITIONAL INFORMATION COVER SHEET I
ADDITIONAL INFORMATION NO: O 3.3.j APPLICABILITY: CA, CP, DC, WC REQUEST: ITS Section 3.3.2, Notes to Required Actions G.1 and H.1.
Comment: The note allowing surveillance testing in Actions G and H allow bypass of a train for any surveillance test. The existing condition [23] only allows bypass of a train ,
specifically for testing per specification 4.3.2.1. This excludes bypass for CTS 4.3.2.2, !
response time testing. This relaxation is neither noted nor justified in the CTS markup.
t FLOG RESPONSE (original): DOC 1-66-LS45 has been developed to justify the inclusion of response time testing during the bypass of an inoperable channel or one other channel for DCPP and CPSES, FLOG RESPONSE (supplement): Based on a telecon with NRC staff on January 20,1999, NSHC LS-45 has been revised.
ATTACHED PAGES:
Attachment 9, CTS 3/4.3 - ITS 3.3 Enclosure 4, page 85 l
i
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- INSERT LS-45 Q 3.3.J t i ;
l l IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS I NSHC LS45 !
l 10 CFR 50.92 EVALUATION !
FOR !
TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE 3 REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS !
This proposed change would delete the reference to Specification 4.3.1.1 and 4.3.2.1 in the !
allowed bypass notes contained in certain CTS ACTION statements, thus permitting any testing needed for restoration of the inoperable channel. The CTS allows certain inoperable channels, or for some plants one other channel, to be bypassed for testing per Specification 4.3.1.1 or 4.3.2.1. The intent of the bypass is to allow testing that will verify that the channel has been restored to an OPERABLE condition and can be retumed to service. The allowed I bypass time for testing has been evaluated for its impact on the safety analyses. It has been :
determined that the impact of the allowed bypasses is minimal. Whether the bypass is for I setpoint verification or response time verification has no impact on the safety analysis. The types of surveillances performed under the auspices of the ITS 3.3 Reauired Action Notes are not limited by WCAP-10271 or its supolements. WCAP-10271 and its supplements made assumptions on the total duration an analoa channel is bvoassed per year for testina. Those l
assumptions on testina in bvoass are not expected to be exceeded by the proposed chance.
! The proposed TS change has been evaluated and it has been determined that it involves no l significant hazards consideration. This determination has been performed in accordance with
! the criteria set forth in 10 CFR 50.92(c) as quoted below:
l "The Commission may make a final determination, pursuant to the procedures in 50.91, l that a proposed amendment to an ' operating license for a facilitylicensed under 50.21(b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facilityin accordance with the proposed amendment would not:
l l 1. Involve a signiHcant increase in the probability or consequences of an accident l previously evaluated; or
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety."
The following evaluation is provided for the three categories of the significant hazards consideration standards: )
i
- 1. Does the change involve a significant increase in the probability or consequences of an I accident previously evaluated? )
i I
- Overall protection system performance will remain within the bounds of the previously
{ performed accident analyses since no hardware changes are proposed. The proposed change ;
allows surveillance testing as required to retum a function to an OPERABLE condition. The proposed change in the allowed testing will not affect any of the analysis es 6 -
+-
IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATION Q 3.3.J l
NSHC LS-45 (continued) assumptions for any of the accidents previously evaluated. The proposed change will not
- a#cci the probability of any event initiators nor wik the proposed change affect the ability of '
- any safety-related equipment to perform its intended function. There will be neither l degradation in the performance of nor an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation. Therefore, the proposed change does not involve a significant increase in the probability or consequences of i
an accident previously evaluated.
i L Does the change create the possibility of a new or different kind of accident from any l accident previously evaluated?
l There are no hardware changes nor are there any changes in the method by which any safety-related plant system performs its safety function. The change in the surveillance requirements
)
l l will not affect the normal method of plant operation. No new accident scenarios, transient l precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not affect the acceptance criteria for any analyzed event. There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on any margin of safety.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS-45" resulting from the conversion to the improved TS format satisfy the NSHC standards of 10 CFR 50.92(c); and accordingly, a NSHC finding is justified.
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. ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.3-55 APPLICABILITY: CA, CP, DC, WC t REQUEST: Revise ITS SR 3.3.1.16 and SR 3.3.2.10 to verify " required" response times, '
accommodating those channels that have no response time requirements per the current licensing basis. [As such, line-item references to these SRs in Tables 3.3.1-1 and 3.3.2-1 can i be deleted. A similar revision to the ITS 3.3.6 SR Note has also been made regarding SR 3.3.6.8.]
i Comment: Reject - The ITS proposes generic changes to the STS by deleting ;
SR 3.3.1.16 for applicable functions in T3.3.1-1 and modifying SR 3.3.1.16 that are not included in an approved TSTF.
Adoption of WCAP-13632 for eliminating selected sensor response time testing i requires staff review independent of the Conversion TS review. !
TSB management sees no benefit to making this change because the proposed STS '
format introduces ambiguity in identifying TS required testing. Additionally, this change is generic and the staff has not received a TSTF request to review these changes. Bill
[Beckner] notes that the staff is very sensitive about changes to response time testing requirements and although the staff approved moving response time lists to owner-controlled documents, elimination of TS required RT testing requires a TS amendment in either the old STS format or in ISTS.
t FLOG RESPONSE (original): Based on meetings with the NRC held between September 15 and September 18,1998, Diablo Canyon will also be taking this change. The Note preceding i the Surveillance Requirements in ITS 3.3.1 and 3.3.2 has been modified and the wording in *
' ITS SR 3.3.1.16 and SR 3.3.2.10 has been revised to cite the licensee controlled document where the responsa time limits are contained in addition, surveillances originally modified under JFD 3.3-31 will be revised to be consistent with the changes to ITS SR 3.3.2.10 since ,
the affected surveillances measure ESF RESPONSE TIMES.
For WCGS, JFD 3.3-55 is now being adopted in its entirety, which also results in their adopting JFD 3.3-31.
FLOG RESPONSE (revised): Based on a telecon with NRC staff on January 12,1999, the FLOG will adopt the STS wording of SR 3.3.1.16 and SR 3.3.2.10 and will emulate that standard wording in any additional response time surveillances adopted as a result of converting the CTS to ITS (e.g., Callaway SR 3.3.5.4 and SR 3.3.6.6 will not reference the FSAR). ITS Tables 3.3.1-1 and 3.3.2-1 will be revised to cite the response time testing surveillance only where it is required by the current licensing basis. Where not required by the current licensing basis, SR 3.3.1.16 or SR 3.3.2.10 will be struck-through, as applicable. The SR 3.3.1.16 and SR 3.3.2.10 Bases have been clarified regarding some Functions (e.g.,
ESFAS Safety injection) for which the licensee controlled document containing the response y .
time limits must be consulted to ascertain testing requirements. For example, not all circuit paths fed from an SIS (e.g., CCW start and turbine trip at Callaway) have response time w .
. . . - . _ . . .- . .. .. . . - ~ . - - - . . - . - -- ..
L testing requirements. Since this is a revised response, the attached pages include all changes from the original response for which NRC approvalis requested.
l i
ATTACHED PAGES:
Attachment 9, CTS 3/4.3 - ITS 3.3 Enclosure 2, pages 3/4 3-1 and 3/4 3-13 Enclosure 3A, page 1 Enclosure SA, pages 3.3-11,3.3-15,3.3-16,3.3-17,3.3-18,3.3-19,3.3-20,3.3-34,3.3-35, 3.3-37, 3.3-38,3.3-39,3.3-40, 3.3-42, INSERT 3.3-42, 3.3-43, 3.3-44,3.3-45,3.3-58, 3.3-62, and 3.3-63 Enclosure SB, pages B 3.3-60, B 3.3-69, B 3.3-70, B 3.3-138, and B 3.3-143 4 Enclosure 6A, pages 4 and 7 Enclosure 68, pages 6 and 9
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, '3/4.~5 J5 ' REACTOR TRIP SYSTEM INSTRUMENTATION $ /-A GEV l .+ tlMITING CONDITION FOR OPERATION ;
3.3.1 As a minimum, the Reactor Trip Sy~ stem instrumentation channels and l
interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3-1.
l ACTION: .
As shown in Table 3.3-1 /_.g i SURVElltANCE REOUTREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and l the automatic trip logic shall be demonstrated OPERABLE by the performance of the. Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3.1. , -ven hed, 4.3.1.2 RIP SYSTEM RESPONSE TIME of each Reactor, trip funct /-d Q-l.G shall beThe d:...gEACTOR i
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..:tr:t9to be within its limit at least once per 18 monthsOch t::t :h:1.a jg y.,4 g Neutrondetebtorsareexemptfromresponsetimetesting.
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- . LIMITING CONDITION FOR OPERATION !
3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLEmith th:f r '-i; 2 -c /-A
"-t;:t-t: ::t copsistent with the gValues shown in th: '-tp htp:1.! ::1=. f Table 3.3-4. g 2-v3-LG-anA A fj,,gI, APPLICABILfTY: As shown in Table 3.3-3. .
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.. _ __a _, - . . . ._,.,__i_ _ L _ _ _ ae
.. A..L .._ ...... . . . . . . . . . . . , -.
e n .,. .t.
, ,___ .u,. ..J . , _ _,...L,_
& 3-o>A @s)
_. . _ . . . . _ _ . . _ _ - - . 6 _ -- _. .L. . a,.,,- .
O
- r ,---- - q--q r r rr'**=___**_'**
_ ____A --a q '.d
. e. e. - . . . . . y.3 _- . 1" AL_ _L_ ,
.. - o. .. . . L 1. . .. . .._.A.f. .g . .. o . . .u. . 1. .
, .....__a .. u _n m ;
.._ . . i. nt a. a. ._
. . . . . . , . . .. nae._ma n e r _. 6. 6.
- .e . i. _ y ._ 2. n_ i..y . ww aak aL. 2_ e.s__2_A ..M...
. . . . d . e. a...
. . . . . . ._ -. v. .
7 ...r- - -====*
r....ss _ e < , n . , - .e n
_ ..,_ ..-. .e.. > n x ., .a en g.pg A
/ N"""* $ ~.).g2 m , .L. . . . , .-_-e,._ , . , ..L,. , , , ,_ .L.
.L... 1 m
- K. o.- .;_"._-___._.2. ..s.._ is_
n
______i ____, ., __t -- _
.. _ . . ..... g."
s.. . t. ..,. . 2 .t.___,
r*'=="= -r="i =* = = = * * ' - '
e,u__ . t. ._....__.ms... ,,. ______. ....s ., )
.e - . . . . . . . ._ _ _ .._. .L. <
...... _- .t. .,.. ,. ___s...,._,.......,e.,_..,_____ .. .. _
ar_L9-_ e_e. a_ ,__
_ . _ . .AL_
. . .. __. ,.,._a.J. ._. .. _
nL ___,
_e. ._. .
._J e
........ .e...., a _,
.n -
, _ , . . .- _ - ......,.._m,.._..,..,..,
,m,_ , ,,
. ,. ns u.
.,...-. m__,
.m s.....m...__...., .. ..... ... ,
4,- With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.. ,n , .rrA ccr n d SURVETLLANCE REOUIREMENTS N #
4.3.2.1 Each ESFAS instrumentation channel and interlock and the aut tic ua on relays shall be demonstrated OPERABLE by the perfo nee of ;
t 111ance Requirements specified in Table .3-2.
Q'd.'3he sSESFAS Inde"--n ation + "verm F ,e ,
4.3.2.2 ERID SAFETY FEATURES RESPONSE TIME of each ESFAS func tion snall be t--St ;td'io be within the limit at least once per 18 monthsif(the gg i
provisions of Specification 4.0.4 are not applicable for the steam turbine-driven I'#N i auxiliary feedwater pump for entry into MODE 3. The ENGINEERED SAFETY FEATURES ;
RESPONSE TIME testing for the steam turbine-driven auxiliary feedwater pump is .
required to be completed c. L . with,in,_24 _,..;_hours ,after attainin,g 1900 psig in.all st,eam 9_ enerators) m . . . . . ... . ., . ..a ... .... ____
. __ . ,......_._t3.. ... ...
___,,___u...
... .... .. . ._, __ ... _.._t
_ _ _ , . . . _L.___,....___....u....,_._
t.._ u. . . . _ . _.
_..a u...._..3.,. .
. . . , c.. - _.. . . . . . . . . . . . . . . . . . .
r.. . . . . . . . . . . . . . . .
v't5_L_..._,....
i
_.._L__ _, _
. . . a... 6. .m .J a. ...
. ...6 ____ ___ .
7.. f a. s. ~_ _ . 1. .e _..
. ___ ..a.. L. . .. .
.L___ M
. 2. A. L. a ._ &.
9
.e , . . _ _ . , _ _ __ _m ,_
_ ,. .. . ..,..n.,.m____,_ . . . . . . . . . . . . .,...... e.e.- . . . . . . . . . . .. . . . . . . . . . . u. . . . . , . . . ,
u.
.. .. .eL____,_.
. . . . . . . . . .e..-.
.. ..,.2,._. .
- i. an. wad S.,. e.d F~unen. l-et-A a:
C MSe~+e Y - UNIT C.2%1 an 3/4 3-13 Amendment No. M4,108 m.su.wo.e.6e.,aw.-euw.^ _ - - -
_ _ _ _ _ . _ _ .__ __..-._-.__._m._.___.__._._ _ ._._._ _ _ _ __
,.,.^ DESCRIPTION OF. CHANGES TO CURRENT TS SECTION 3/4.3
. This enclosure contains a brief description / justification for each marked-up change l to the current Technical Specifications. The changes are identified by change
( numbers contained in enclosure'2-(Mark up of the current Technical Specifications).
-In addition,' the referenced No Significant Hazards Considerations (NSHCs) are contained in enclosure 4. Only technical' changes are discussed; administrative ,
changes .(i.e., format, presentation, and editorial changes) made to conform to i NUREG 1431 Revision 1 are not discussed. For Enclosures 3A. 3B, 4, 6A, and 68, text i in brackets "[ ]" indicates the information is plant specific and is not common to l all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that !
- other JLS plants may have plant specific information in that. location.
l CHANGE NUPSER R$K DESCRIPTION j
- 1 01 A A Note, " Separate Condition entry is allowed for each ;
Function " is added to the ACTIONS for the Reactor Trip l System ESFAS, [ Remote Shutdown, including each required )
, ASP control], and Accident Monitoring Instrumentation.
This change clarifies those situations where the current TS ACTION Statements are not uniquely associated with a j l particular Function or where the required channels are i I .g- specified on a per steam line, per loop, per SG, per bus, l ,TZM etc., basis. This change is consistent with current i l v' operating practices and NUREG 1431 Rev.1. [ Relaxations )
associated with this change are discussed in CN 2 46 LS 42.3 :ZN.fEXT~3A-l $ l~'Ol'"A l
1-02 LG The current TS require that response time testing be d
[ performed on each reactor trip and ESFAS function everyDC.-
18)nthsandthatalternatetrainsbetestedin ;
L successive tests. The current TS description of the -
channel testing protocol ' matches the improved TS i definition of STAGGERED TEST BASIC. However, several trip ;
functions do not require response time testing, as !
'ndicated by N.A. in the tables of rauwise time lim _ite .
The improved TS specify that @ response time [previously r ,
testing be performed on a STAGGERED TEST BASIS and do not j impose any requirements as to which train should be ;
tested. Therefore, S 7.3~63~ ;
L l
~
ps!FrfEs&frB@Pthe
, 1 03 LS l ' In current TS SR 4.3.1.2 and 4.3.2.2, the active verb is .
j changed from " demonstrated" to " verified". This allows !
! Reactor Trip System and ESFAS sensor response time j verifications to be performed per WCAP 13632 P A ,
i L
Revision 2. " Elimination of Pressure Sensor Response Time !
h DESCRIPTION OF CHANGES TO CURRENT TS 1 5/15/97 ,
~
p in..o -.-wa: a-wa ,.y g-- + m- **
l RTS Instrumentation l
3.3.1 SURVEILLANCE REQUIREMENTS
'.' - LQ 2.3-55 t rs ha .,"--}
........................................N0TE- -- -
Refer to Table 3.3.11 to determine which SR 7 Eg g gjcp,Lapply for each RTS Function.
r .1_
,h 1
/
t 1
.' 1 ' . .'3 l . ^ . . . . . . . . . . . . . a 0. . . I ff. _
SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.1.2 ------ -
- NOTES - -- - -
- 1. Adjust NIS channel if absolute difference is > 2%.
- 2. Not required to be performed until -12 24 3.3 % -
hours after THERHAL POWER is 215% RTP.
P
.i Compare results of calorimetric heat balance 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> V calculation to Nuclear Instrumentation System i
(NIS) channel output.
j SR 3.3.1.3 -- -- - -- ---- NOTES--- --- -- - -- i
- 1. Adjust NIS channel if absolute difference is a 64- 2U 3.3 2. Not required to be performed until ;
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERHAL POWER is B 2 M 4 S0* RTP. 3.3 96 Compare results of the incore detector 31 effective measurements to NIS AFD. full power days (EFPD) i
! (continued) l i
r .* a
':,~
i HARKUP OF WOG STS REV 1 (NUREG 1431) 3.3-11 5/15/97 l
l
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 4
SR 3.3.1.12 Nottused. *
-3.3-50 NOX This Survcillance shall include verification of
.ncacter Occlent Systa resisten;c t a grature i,ctcctor bpass 1;;- 'l;w retc.
cr Perfora 0".'?!IL CALIOPJTION. 10 eenths SR 3.3.1.13 Perform COT. 18 months i
l SR 3.3.1.14 -- - - --- - -NOTE -- - ----- - --
l Verification of setpoint is not required, j Perform TADOT. 18 months B ym- -
)
SR 3.3.1.15 -- ------- - - NOTE -- -- -- - - -- - - NOTE- d-
- j Verification of'setpoint is not required. On requir i
............................................... when t perfo within i previ s 1 days Perform TADOT. P ior to reac r tar UN !
u & 3M SR 3.3.1.16 -- ---- - - NOTE -- -- - --- -
Neutron detectors are excluded from response i i
time testing. / l A---
i Verify 6 TS RESPONSE TIMES'are 4s 18 months on a B- l; within 11 mig VMo46 649E STAGGERED TEST gpf,gy) 2
~Q -
hrier-lr exceedi O c 0 1 in4tebch
- v MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 15 '" ""*"M ' * -/- b /cen 5/15/97 W3
-sep-- oe- mw+.A4-9 frevrona} i nof fer-Armedinfe.
1 I
RTS Instrumentation j
, 3.3.1 l
Table 3.3.1 1 (page 1 of 9)
Reactor Trip System Instnnentation j i
APPLICABLE MODES j OR OTHER SPECIFIED REQUIRED . SURVEILLANCE ALLOWABLE Htte 3.3 08- l FUNCTION CONDITIONS CHANNELS CONDITIONS ' REQUIREMENTS VALUE" GE#GIM* ,
)
l
- 1. Manual Reactor 1,2 2 B SR 3.3.1.14 NA e i Trip l 3*,4*,5* 2 C SR.3.3.1.14 NA M i
- 2. Power Range Woutron Flux l
1
- a. High 1,2 4 D SR 3;3.1.1 d 12;31 RTP t-4996-RfP B-PS-SR 3.3.1.2 +tt:0
~'
C=
S 1
. 3 . " " .3 g --
- b. Low IIC) 2
, 4 E SR 3.3.1.1 h'J8:51RTP . = ::: B PS-SR 3.3.1.8 ihh2 h,f-6 f J MNO SR 3.3.1.11 l
^^
rfn-ke-44sy4 O';;' O V M 6 W) b@ h u. Power Range ^
, " . Neutron Flux Rate g ;
i l w; High Positive - 1,2 4 E 'SR 3.3.1.7 k%:56:6 T-55-Ate I Rate SR 3.3.1.11 % RTP .....;- !
with time cenatent B-PS ;
constant t-0-see a 2 sec L L. .g. -. .... W 4 e = :.:. .7 a :,.= ;;;; se**P 3.3-014 4 i Ret. = :.:.:.;1 .:..,,- . . . . . l l = 3.: . '. . "i, conetont constant h eh
- 4. Intermediate Range 1 *, 2"" 2 F ,G - SR 3.3.1.1 s'l55:3E . I',"; ;;;r
- g.p3 3 i
Woutron Flux SR 3.3.1.F. 59E RTP i l SR 3.3.1.11 ;
- 1 I e* e * *R4rrr e 3.3 95. l 91Hbb
- t.4
= :.:.; ;; i i
I (continued)
(a) ::. . ; .- . . ;..,;. . a . ; . . . n . : ._ ;. . . . -, .. . . . . . _ ,; , u ; ; . . . . ; _ a.s, . . . _ . : ;, ; . .. :; ;,
-.:..;.,; , ... L, .:.. _.; . ThemitouablevaluedefinesxheyJoitinsisafety,. system sett4nsesse. shew 3.3-08:
for.:the~~JtM5e* points.
...J Rod Control System capable of rod withdrawal o
'i not p3.3.-m/s j (b) With :...i.,. T ,,, ;;. ..;.. . (":77,.; d tullyj. inserted.
..J
- 3.3 122 l (c) Below the P-10 (Power Range Neutron Flux) interlocks. MC SP*1Plfr'$ i (d) Above the P-6 (Intermediate Range Neutron Flux) interlocks. i c.; :;.;- .: : ;;..;.. ;.... :. ;.......;-; ...;..; a ..
3.3 95-i MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 16 5/15/97
RTS Instrumentation 3.3.1 Table 3.3.1-1 (pege 2 of 9) .
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER 3.3-08'-
- SPECIFIED REQUIRED SIRVEILLANCE - ALLOWABLE 9*te FUNCTION CONDITIONS CHANNELS- CONDITitJNS REQUIREMENTS VALUE"' GE4P&l#T* f f
i 1,J s-+ + t5 3.3 41 >
- 5. Source Range ' 2*' 2 SR 3.3.1.1 s:;12 4d O'
Neutron Flux q g,yg SR 3.3.1.8 E5 cps cpe Sa !..!..M. . .5 3.3 55 i 3 *', 4 *', 5 *' 2 J,K SR 3.3.1.1 101M 4d . ;.: c .3.3 41- i SR 3.3.1.7 E5 cps epe g-SR 3.3.1.11
= :.:. . = ,3.3-55+ .
a*r-4*rs * = :.:.;.;
4 t-
- 3.3 41 ' -
1 I
- 6. Overtemperature AT 1,2 6 E SR 3.3.1.1 Refer to Refer-to g SR 3.3.1.3 Note 1 # wee-4 SR 3.3.1.6 (Page tPope se 1.117 3.3 25) F.+-ett 5hr V '8CB832 ^ -
3.3 50 : i E 3'5'- ED
.. [, 8. M. -- a h 5 S'"5 .- u
- 7. Overpower AT 1,2 5 E SR 3.3.1.1 Refer to Refer-to g-SR 3.3.1.7 Note 2 Note-2 M (Page tes,e 3.3 50:
- :.:." : 3.3-25)' 5+5t '
'.'.'.=3 ...,
(continued)
(a) :. . . ... . . ;.. ; . . .... . ..:.. . :_ _...;.. _ , . ...;.. ..;,;.:. ,;. ..: ,
. . . . . :. . . . . n :, .,3.3 08 4
_;.._: , ,, .:.. ..: .. pw3nnehabww luwweth:thettalths m e]ments 1 forJ$heNrAimstpoint*~.
(b) With Rits eteced-end Rod Control System capable of rod withdrawal SL 3 hrIN! 3*3 122' (e) Below the P-6 (Intermediate Range Neutron Flux) interlocks. m e,. m yg '3.M l
-+*- u . . . u.e : ;. . . ..
.......y..,..
_ _ [. u . . . :. 7 ..g . . ;.
- . M ..-t .
3.3 41:
.. . .. .... -,... ,....... ..... . . .r.. - ,, ,.. .,,, ., .. ..... . . . _ , . _ _ _ _ _ - _ _ __.
,_ _- - ,m --
M 7
MARKUP OF WOG STS REV 1 (NUREG-1431) 3.3 17 5/15/97
RTS Instrumentation
_ 3.3.1 Table 3.3.1 1 (page 3 of 9)
Reactor Trip System Instrumentation APPLICABLE MODES oR CTHER -
SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE Mfe 3.3 08-FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE" GE4POWi* ,
r
- 8. Pressurizer Pressure ,
- s. Low 15' 5 M SR 3.3.1.1 4666 i ^00 v.,, .B PS SR 3.3.1.7 53874 psig --
C : !.' ' ")
p.s-ss (Ach n~vc th-o ke-kmL
- b. High 1,2 4 Y SR 3.3.1.1 2596 . ::0; r. 4 g.p3 SR 3.3.1.7 IC3400psig
.w c5!:!:^$3 i
- - -w
- 9. Pressurizer Water 1 *' 3 M SR 3.3.1.1 I"ip,UEE*4f t-92M g.p3 Level - High SR 3.3.1.7 lpetesment SR 3.3.1.10 apan peq:. Reactor Coolant 38 Synet leap N BR35MM EW' ..
3.3-42.
r-? . [ Flow - Low BR333BN 09:e ens e an - _
~
.pt 3.1. l. Ib G 3.3-5.S ,
.. ::,,;e L ,, 4* 5 per M = :.:.1.1 e%s e-9e" '3.3 % ?i+:'*i. :
r :*: ' :,
1
- . . - L . ,,. 4'" 5 pee M = :.:.1.1 ehe r-9e" :3.3 42
% %ii+i.
= : *: * : * : ~,
(continued)
(a) ;.... .. . . ,i ; ei . , f . wL. ~. a .; .
i;,,,J,,;.,,
- ..,1..
.,J ;,,
, ,e ,_ , ..,, a . ,,, a , ;d ; L L ^.'. ".
. ,;;. TheM11osibleNaluejdefines3he3faltingsfetysvstamisettihig3ee;$he;5ases J., . .J , . . ,,, , Li,~ , , ; ; r
'3.3 08 fotfthe2c1p7,etpolots; (g) Above the P 7 (Low Power Reactor Trips Block) interlock.
^;, j ;,L.,..
q., - ^ "; ;.< ,j :.. i. . T; - { ,,,ig,-; ~" ,, ~. . ~ - . . , . .. . . . , , . . 3.3-42
,,-.7..., ,
, , , , . . . . ~ . , , . . . . . ,-
.....,.m.
Ia) KFof?loopmilrGassiimaesuriedfilow,tMMW1PJf660 som);
3.3 09 MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.3 18 5/15/97
.O-Wt a
RTS Instrumentation 3.3.1 Table 3.3.1 1 (page 4 of 9)
P. rector. Trip System Instrumentation ,
i APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE 9tte 3.3 08 FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE" GE4P M M " ,
i 11- fd W__ ..
-3.3 ."'".,","..i"",".'".".'
Poeitten 4" 4 per 0 7.: *,.3. ', . % m M i
.. ::.,;; L ,, nee '
4* + per M :: :.:.; a a m j
- . . ; .,., eep ;
l
- 12. undervottese 1* 5 per M SR 3.3.1 9 a pouslerec '
2 -, :: ;..
/ B PS t RCPs h
~7 5RC 3.3.1.1!.;.1 LO 4760
/
- h h3.3--W j MVChk'thMf
- 13. Underfrequency 1* 5 per M SR 3.3.1.9 a 5L1 Hz i 57.5 ;;'
B PS' i RCPs tums SR_ 3.3.1.1 g;.s. -
t TJann
-m n;~ ;.;: ' 3 Q
l u
to neienteeminirtor :3.3 tamanner:ME Lemmaus)
- r. sedisessuunsfer L132 User 3sp 5 msnmes s1paangy 3.3-M -
luendisemp1 m3ENN
==*="a lismew testnum ~ ' -
- (SK?.2./.14 f13.1-SC emuseuiist ase mesuseep
- 10 sticionirsihigiit U2 Figrlso g assm'i k W asler 3.3 ussociuses massaw memsw inweeammadiirt n;smo mense n=58eaat JA 7.3./. //, 8=*a'si" 87.76 sadnamano men Esistishiid)
(e) ;; . . . . . . ;., a . u. . . . .n . . . . ; . . . . .: . . _ , . . . . . _ .; , :. ; ; ; :,; . ;. ; - . : . . _ . :. .,, ,n . . :; :,
i a:. ;.,, : ;,, :. .;;. theAtt%le!weisefdefjnesitheMinisins safety systemwtinsmseejehehiess 3.3 08-i Fan $heNelp Setpoints; I (g) Above the P-7 (Low Power Reactor Trips Block) interlock.
i
- .; ;.; .. .
- ..
- t r- . :.. . ;... . . . . ; -; ..;..; a .
I , n ; ,; ,.. .:.. t 7 ;;a r-- :...... :. , ::a; ...a.- ;a .; : ;, ;- .:.c r : ;r-, :.... ;. ; . .. :s; 3.3 01-EU tisienspifestil@lDDEsfferl%h$ennels#tCTable;332f1]are1mor@ipstrictlyii. . 3.3-52
- G J
MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 19 5/15/97 my, ,-.--,mem
. __- . _. _ _ _ _ _ , . . _ _ _m._ _._.__ ._- _ _ ..___ _ . . _ . . . _ - . . _ _
r RTS Instrumentation !
3.3.1 i
1
, Table 3.3.1-1 (page 5 of 9)
Reactor Trip System Instrtamentation
- A M ICABLE MODES
.2 OTHER , ' 3;3-08=' i SPECIFIED REQUIRED SURVE!LLANCE ALLOWABLE tete I
FUNCTION CONDIT!DNS CHANNELS CONDITIONS REQUIREMENTS VALUE" GE4P91*T*
t rogsequsiis) t r
- vessWtilta IU2 4- W sRs
- 3m '3.3-046
_ some.neii e 3R 5. n t:7 tr3ashin st373T3D - ?
una p s:r t.it Q1755' .
stessiiinse- icve.sairn cA-7.3-de 7 l
+ ~ ~ Equivalent i
^
go35 igg 3. 3**lW l aTra j irwisselsa C/$-7.3-do*7 Equimum at i s togs:ex 7.7-/fd ;
, RR" encarans can l gp y ict:ogiig8
- - t" . d. sintetiiisist a sessnti "3'.3 04T~
~
1 posesuman- sR53-rI b' Eq essous an JYs.s.t. /(, 'i Q***** & zs-sS~
- 15. settuiip +;e e per-se e ; :.:."." t-se:*m s e 23.3-01r te1 ster ; !.!.".7 . i i.... ;- Z ."."..^0 42 e . 4 ."; l ;; 1 0 f.;;
eetnetdorrMrvetr 7 t pe,-ee ; ! .: . '. . ". ;
steamr-f+ourf ; .0.. 7 stesur+ ten steamr-f4ew fosenster-f+ew ; :.3.'."; seit9e stit9P tevenmech ; !. .".";
- S6. Turbine Trip ,
- s. Low Fluid Oil lu) 3 . a-g SR 3.3.1.10 956 . 0; ,.- ,; g.p3 Pressure O 27 02 '" 3 7,7 -d2 3
- b. Turbine Stop tu) 4 P SR 3.3.1.10 2 1% open t-ilt y .B__
valve Closure SR 3.3.1.15 (continued) r (e3TheMitsumb$strelseedefifies::theWiisilti$sisefety s'ystem mettire.;Mthe;Beses3for kthe.7rJp:Setp61nts. 3.3 (j)Above the P-9 (Power Range Neutron Flux) interlock.
ftiWithm -tineidelay 42&D sseconds, 3.3 04 wwith'.exime*tay;s.;ssomconds. ,
11 MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 20 5/15/97
- 5
-- :,- 4
~
.1 i ~
ESFAS Instrumentation
.- 3.3.2 y ,_
SURVEILLANCE REQUIREMENTS 7#
.......................................-NOTE - - - -- - -- - --
gy' [M Rafar to Table 3.3.21 to determine which SRs ?oons_erjt erggp~bpplyforeachESFASFunction. -
, .s .an$jeti ;ica
.1
---sg 1 g:. gQ y 9 55-
... .. . ... . ............. ..... ..... .. .. g 7.. .. .
/ ' /V U SURVEILLANCE FRE0VENCY SR 3.3.2.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR S.3.2.2 Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TES BASIS cg_y,3.,9 SR 3.3.2.3 - - --- ----- ------NOTE - - - -- -- - -
The continuity check may be excluded.
)$yg
.er ...............................................
Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS
\
SR 3.3.2.4 Perform MASTER RELAY TEST. 31 days on a STAGGERED TEST BASIS l
l SR 3.3.2.5 Perform COT. 92 days SR 3.3.2.6 Perform SLAVE RELAY TEST. 92 days B l
f w (continued)
! --ye7s ---
~'1 No + ,gire Lle h .rh ve rel ,
, .x K40.1, K426 K6.22, K62+, Kik & 3. 2-/9
'?
M7+0) and K1+/. -
6 MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.3 34 5/15/97
ESFAS Instrumentation
, 3.3.2 SURVEILLANCE REQUIREMENTS (continued)-.
SURVEILLANCE FREQUENCY SR 3.3.2.7 =
"crification of relay Octp;ints not requirc-d. 3.3-136-Perform TADOT. $192 days. B PS SR 3.3.2.8 -
-- -- - NOTE - -- -- - -
Verification of setpoint not required for manual initiation functions.
Perform TADOT. 18 months B --
SR 3.3.2.9 - -- - --------
--NOTE - --- ---- --- -- --
'O This Surveillance shall include verification
.. that the time constants are adjusted to the prescribed values.
Perform CHANNEL CALIBRATION. l8 months B .-
SR 3.3.2.10 - - - --
NOTE-- - --- - -- ---
Not required to be performed for the turbine B driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SG pressure is 2 -1999 90_0 psig. _ B PS-Verifysedssp1ES RESPONSE TIMES are within 18 months on a M limi W #gM, STAGGERED TEST B BASIS Q.r.s-sc ,,,___
(continue MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3-35 5/15/97
_ _m. - _ _ . _ _ - _ _ - _
ESFAS Instrumentation 3.3.2 Table 3.3.2 1 (page 1 of 10)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE N00ES OR i omER SPECIFIED REQUIRED 3.3 08 SURVEILLANCE ALLOWABLE TRH' FUN 0 TION CONDITIONS CHAndLS COM)IT10NS REQUIREENTS VALUE5 GEPOIW"
- 1. .Safetyinjection
- a. Manual Initiation 1.2.3.4 2 B SR 3.3.2.8 E M
- b. Automatic Actuation 1.2.3.4 2 trains C SR 3.3.2.2 NA M Logic and Actuation SR 3.3.2.4 PS-Relay? [$$PS) {
SR 3.3.2. 1 Sit 3.3.2.13 0*I.3-M s c.. Containment 1.2.3 3 D SR 3.3.2.1
- 14;5';psig s . 2.0 pi; .g.ps Pressure - High 1 SR 3.3.2.5 B-66 SR 3.3.f'9 -
'iHHHO
- d. Pressurizer 1.2.3* 3$ D SR 3.3.2.1 P2834Jsig . ;02 p9 .g.p3 Pressure - SR 3.3.2.5 1839 Y *** V# $
e.
&'E.'3-W ( Yrf e-W f Steam Line Pressure '
,agts
'M e Low 1.2.3* 3 per D SR 3.3.2.1 E5713s1#8 r-675 B-PS l
_ steam SR 3.3.2.5 635 ps41l*
line SR 3.3.2.9
-.,.0.2.1D
~
G H hgh 143 B-pee 9 2 .3.2.1 .100 pi; a 07 pi; 3,3 01 Dif fere;.tiel steam 0" 2.2.2.0 Pressure We Z .3.2.0 Rt nn St;; 0" 2.3.2.10 s Ekes f 'it. 0;;; T1;.; "- 4r223
- iHer B 5" 3.2.2.1 M M 3,3 01-T.;; Stc ; Lire; steen 0" 3.2.2.0 he = . . .0
= .3.2.10 0;;rcin..; itt. lyer3* i-pee 9 SR-3-3-i 1 a "0.0*i e-5532F T,., - tow-tw hop = . .2.5 3.3 0" 3.2.2.0
= :.0.2.10 (continued)
(a) ra :.;r stc. un't egc.1x usi. ..;; tun; ny eeaan :n y Aih.au: :::a: 1 :;n; en RtNM St dj ntredelegj uscd by it; ;. nit 3.3-08 y system:settingC5ee the3ases:for the' Trip $etpoints The%llowable Walue definesW!1talting Me F, e :safet_b l
.y,y.4 3 (b) Above the P.11 (Pressurizer Pressure) interlock [anM $8/ev [-l) gn je.rs [delc/, g y,7g g (c) Time constants used in the lead / lag controller are 13 a 50 seconds and t, s 5 seconds.
M *i:n tM a !? (T.,,- L;.; L;.;; Meri;;k.
it; Lcs; tren er ;c, I t; :. f retien (.ifired se er ee ci .,~,.d;ns te M full stca s.; bcle 200 leed. ;. .d LP 3.3 01 wereesy.; li aerly fra ?t: full stes fle et 200 hed te 1140 fu11 ;tce fh.; et 1000 bed end e" ee; ;pr.d,n; to 114 fu ;t;; ik.; :.L i.; 100 '.xd.
(f) Lea.1 then er ;c el te e f retien defin;d es or ee .se;~i.d; .; t; 400 ful' stce; ;.; bet.e;n 0: r,nd 20 hed 3.3 01 O-4,@
- nd ttan a c" _] cre-e;ing 'irer.r'v ffs 400 ;tte; ; ;t 200 he,d to 1100 ful' ;tes 0;.; et 1000 hed.
eri i s hn 6 4~ (N f
.re s-!
g,u, MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.3 37 5/15/97
ESFAS Znstrumentation 3.3.2
~
l Table 3.3.21 (page 2 of 10)
- Engineered Safety Feature Actuation System Instrumentation APPLICABLE l HDDES OR OTHER SPECIFIED 3.3 08 REQUIRED SURVEILLANCE ALLOWABLE TREP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE8" 55TP01E* j l
- 1. Safety In,)ection (cont.) i
- . "igh ",t; n "l w n b-&-4* G-pee B "" 2.0.".1 fet H+ 3.3-01 T.; ",t; a Lire; eteen "" 3 . 3. 2. ",
en kne , ,
- . , . . . e.m, en ,
- . ....e.
0;incit~..; .;itt, h-A-3* t-tee B 5" 3.2.".1 e-635*-psy 2 07; pis 3.3 01
- ^-- '- "-------
stese 0" 2.0.".5 1
-tow kne 0" 2.,,,3 . " . ", ,, l
.n- .... ...
- a. Manual Initiation 1.2.3.4 2 per B SR 3.3.2.8 E M I train, 2 trains l gg b. Automatic Actuation 1,2,3,4 2 trains C SR 3.3.2.2 NA E
' <. F- ! Logic and Actuation SR 3.3.2.4 p3
, Relays [SSPS) SR 3.3.2.6 j
- c. Contairment 1,2,3 4 E SR 3.3.2.1 *s '283Asig s-le-95 Pressure SR 3.3.2.5 '
10-31 ps4g igh - 3 SR 3.3.2.9 (rM]I-wevetk'ilfe-$j
- ; : "3 Q 3.3--SC M,h - a 'T.; Lc;p hers B-eets F " 3.0.0.1 s-le-si s-se-es 3,3 01:
I Plente) eH " :. .2.5 peis peg
-....m.,
em ,,,,
- .,. ....o.
A J (continued)
(a) ra.-.; ;r'; ,:;. Uni: .g ;ifi; w '. ..;;;ica; ;.;y ;;,.::i, only s' n ci; "; M ,~.. :n; an "atp e t 3.3 08-0:.dy ;;r. A n;y ;;d by tr.; ='t The allowable;ralaeasettnes M1miting safetsystes:settswrcsee m,
s t'he_m.... >.m.,
Joro the.* ,, , , , Trip N._ m,
. m. .. ,..m .
.,,__,_-._m.._..-...~.2 > _ ~ _- __ m f m,
. .. ...m._. __ . . . - . , .
, . . . . ..m. , - . . ._. , .._
, , . ~ .... , ._ .... .. ,..,,
.. _ __ ,,_ . ,m,.,.__ m.,. ,__2 __2 ..
. . , . . . . . ~ - -
3.3-01
.-..._._.m..,_
..~
,___.,.,.m.,.,.
, ~_. ,__2..,..._,,.,...,,___,,__,...,.,__2..
. . , . . . ~
__m ..
m.
.., . ..,,,.,. ,..,,. ... .__,,_..m..- .- .- .
,__2 H+-6_..m.. _ _ .__r. .., ., ._. ,_.-
,___m,. _
m,
..~.2 .
__ .. m__
. . , ,. . ..,,_____,,_.m,_.__. __2 m.,. . . ~ m.
. , , , .. .. . . ~ . -. m. . . . . . , , , . . . , .. .. . .. .
- .:; a fic.; ;; ::: b;, t; ::0: fai' ;;; a fu.; :: 100:
. . - ~ .
3.3 01 ac;d ;-/ tun ; or .nereetw; lin;;rly '.
40t+.-
i l@
l MARKUP 0F WOG STS REV 1 (NUREG-1431) 3.3-38 5/15/97
ESFAS Instrumentation
_ 3.3.2 i Table 3.3.2 1 (page 3 of 10) !
Engineered Safety Feature Actuation System Instrumentation APPLICABLI MODES OR l
OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP
~ 3.3 08 i i
FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE'" SRPP" T" !
- 3. Containment Isolation
- a. Phase A Isolation 1
(1) Manual 1.2.3.4 2 B SR 3.3.2.8 NA E Initiation (2) Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA E Actuation Logic SR 3.3.2.4 7 and Actuation SR 3.3.2. 5 p3^
Re1ays (SSPS) gg 1,7.2,/3 h 3'.3-/f (3) Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.
- b. Phase B Isolation T.~i, - .
s ., (1) Manual 1.2.3.4 2 per D SR 3.3.2.8 E m Initiation train. 2 trains i
(2) Automatic 1,2,3.4 2 trains C SR 3.3.2.2 m M Actu.. ion Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 p3 Relays [SSPS)
I (3) Containment Pressure l l
High - 3 1.2.3 i E SR 3.3.2.1 iG28:3jpsk r-19-05 B PS- i C'@ #' Ak_. _
SR 2 _
M- %
eg.g-sr x
%+; n~a Ma-Lgw=.=o
-~- _
s- - - '
l l
- 4. ' Steam Line Isolation
- a. Manual Initiation 1.2"' . 3"' 2 F SR 3.3.2.8 NA E (continued)
I (a) .t cu ; ; etc. Un t +. f;c ;;;n-..;st;;n: 2., ;;c.;;t ;- y Alu.;tu ';;b; :.r. :;n; a r,etw t 3.3-08 !
Study .ct.'.edeksy ;;d by it.; un't The AllowableWalueDefinesithellaiting: safety;systensetting1See i theSasesfoDthe Wrjp$etpoints; (i) Except when all MSIVs are ci m ad r_4 ." =' ' ^ " B-PS Jh IET t ays;- 25 .3 .'K63 ._K740" . I'd SM ebe.t ed- y;38 r: gg 4- -
tr 211 ~ $_ .# 15 f61 y24 ~ s. 1 1 tot' yt ,the tous . days.
Qn-19 MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 39 5/15/97
ESFAS Instrumentation 3.3.2 Table 3.3.21 (page 4 of 10)
Engineered Safety Feature Actuation System Instrumentation 1
APPLICABLI HDDES OR omER 3.3-08~
SPECIFIED REQUIRED SURVEILLANCE ALLalABLE TRIP FUNCTION CONDITIDNS CHANNELS COM)ITIONS REQUIREMENTS VALUE" SETPO4 W "
- 4. Steam Line Isolation fcontinued)
J
- b. Automatic Actuation 1. 2"' . 3"' 2 trains G SR 3.3.2.2 NA M _ p3 _
Logic and Actuation SR 3.3.2.4 ,
Relays [SSPS) SR 3.3.2.6 '
l p Automatic; Actuation lLn2*;3* 2 p 4amese:2r- NA M--23@g 3.3-98 l LogicDnd Actuation trains" .1% f.1,2,2 i Relays 3i(NSF .5) p,' e- Coi.tainment 1.2"' 3"' 4$ D SR 3.3.2.1 iG18'5]pf.g 1 r+35 Pressure- High 2 SR 3.3.2_.5 6-61 ps4g i N SR 3.3.2.9 N C 2.2.2.O f) E.3-55 J'-lch ruwe Q
-.. f ti & Stean Li Pressure &ktkreuf] )
(1) Low 1.2 "' . 3 per D SR 3.3.2.1 kJ571;psig" t-95 B PS - ,
line C.T 3.0 23 !
3*""
(2) Negative 3 per D SR 3.3.2.1 M248%'Rs1 r-110 g.p3 Rate - High steam SR 3.3.2.5 ish6 ps4/+ec* i l line SR 3.3.2.9 pet /see j M ;.3.2.! D
^ N _
.7
- . "i p St;;; "l;.; "- T.;; h2* -3* 2-per 0 0",0.0.2.1 iet (f) 3.3 01>
Stc;; Line; steen S" 3.2.2.0 44ne 0" 3.3.2.0 SR 3.2.2.10 l 0;inci& nt m th h2*; i-per 0 0" 3.2.2.1 . 000.0 7 t-5532f T, - tow-tow 3*"' icep 0" 3.2.2.0 SR-4-3-0-9 l 0" 3.3.2.10 l (continuf4; (a) *:c.i;.;;r'; Stc. Uni, ;g;;fi; i;+1a..t;ti;n; ny ee,.t;i ; enly All; ;ti; V;1a; igr.dini; en Sctpint 3.3 08 l
Study rctt.; del;;;y and by th unit The 413 cwableWalue #finesithe)1miting se systes settingG5ee thelases for th:eMp $etpoints; v4e Arse en S.743 i
(b) Above the P-11 (Pressurizer Pressure) Interloc[or/[e/ew h// un/c.c/ /sc/cc (c) Time constants used in the lead / lag controller are ti a 50 seconds and te s 5 seconds.
f)?.f43 B I
(9) Below the P.11 (Pressurizer Pressure) InterlockDhaweverimefbejblockedfelow P.'11'when3afety3rdection 3.3 63 on 4ow~steampine pressure!.1sinet'$1ocked;
~
-4 .-
d (h) Time constant utilized in the rate / leg controller is e i;50 seconds. 3.3 12
~
(1) Except when all MSIVs are closed. end c ntiated. B PS
( lo) Eachtr_ainyequires 4 ainimum ofetwo;pr_ogrammab10ogic:controllerstote DPtknC1 3.3-98:
~
{
l MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 40 5/15/97 l
l ESFAS Instrumentation 3.3.2 4
Table 3.3.21 (page 6 of 10)
( Engineered Safety Feature Actuation System Instrumentation i
i APPLICABLE MODES OR OTHER 3.3-08' SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE Mie FUNCTION CO@lTIONS CHANNELS CONDITIONS REQUIREENTS VALUE" SETPOIE" 4 Ste s Lin; bc htica
-. in ;d P "i p, "i p, 0;; ; T h. 172 " , 4 " @ 0 SR-3-3rLM > 100t of . of fail eteen Z 0.0.0.5 Si' :t;w h 49e Z 0.0.0.0 ;.; e,t ti' f+ew-et 3"3 01 i
= 0.0.0.10 ind :t;a ftell l pressiee ind ;t;s l
presstere 1 l
0; ire:1... .;;t', R f;ty n;f;r to i reti = 1 '%f;ty ins;^.icn; fer ;i' '7itieti = f n; tic; 3.3-01 i kj;; tion ad 7;;;;, - ...;.
1 5, l Turbine Trip and , -
7 I'~ [
i Feedvater Isolation l e,
57' M a. Automatic Actuation h[.rrth
- 1. 2* .78' 2 trains M 6 SR 3.3.2.2 NA E Cb~3. 00 3 i g.p3 Logic and Actuation SR 3.3.2.4 . p3 Relays [SSPS) SR 3.3.2.
,rgy, y,::2. H- $ 7. 9-/9 3.3 19 :
f3 C/- 3 b'. Mtematicactustion Lopcnnametustion 1r2*>' 3 painsaa E -+- G,,rpw3.7.2,2 m CA-ZF-g4 ~3.3 98 ReinysgMSgp .
- p. b- SG Water 1. 2* . 3" B 5 per a9 SR 3.3.2.1 D924 s-62-44 :g.p3 Level - High SG SR 3.3.2.5 llarmt; Range High (P 14) - ~ .3.2.9 itistrament g,3Lf'f" hth r-egyg G ; . M . !D Span
.rkke-L y tL er Safety Refer to Func ,)ection$ all initiation functions
. Injection and requirements.
.1E.ir-/+/ (C "ti""'d) xN rd)er .7,342 C A-7. 7-42 (a) n;. a u'; etc. "n't we'i; igt ..t;tia; ny _..;i, u.ly A't :.tk nl.a 4..d:n; = Rts,i t 3.3 08-Stadj n t,' i b;y n;d by tre un-t The Allowable'ValuesleffnesMhe*11sittag; safety; system".settingC3ee the Sases70t-the'Trjp~Setpoints' .
(j) Except when all MFIVs MFRW. ad n;;;;;t;d t,,p;; wh;; are closed, ad d; nt'ated g ucht;d by ;
- u;;d x = 1 n k;. 3*3 64 I ro) Eachtrain'reauires ninimum of waregrammable?logictcontf613ers'ineWERABLE' 3.3 98
- . p) mpt : ay_ dd 11' G as^ % - ;' pr;ior ~' ng-
,, has^ j1n J .6ffor J 4- sf2if t thi he ous tia[I F
- t. - 'theAniP
/ /3 349
- v." ..
/ /_/ /_ \
h$.$~ \ MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.3 42 5/15/97
INSERT 33-42 g y .y ,_ g i APPLICABLE 4 i MODES OR 'l OTHER SPECIFIED ' REQUIRED ' SURVEILLANCE ALLOWABLE FUNCTION' CONDITIONS CHANNELS CONDITIONS REQUIREMENTS - VALUE i 5.e Steam Generator Water Level Low-Lo# (1) Steam Generator 1,20,38 4 per SG D SR 3.3.2.1 218.4% of Water Level Low-Low SR 3.3.2.5 Narrow Range (Adverse Containment - SR 3.3.2.9 Instrument - Environment)- y 9,y,2.jp Span $ 7.7-65~
-N -
(2) Steam Generator 1, 20, 30 4 per SG N SR 3.3.2.1 213.0% of Water Level Low-Low SR 3.3.2.5 Narrow Range (Normal Containment 3.3 2 9 Instrument
, Environmese) Jgy,f,2,/0 Span S 7.1- 55' (3) Vessel AT 1, 2 0 4 M SR 3.3.2.1 Equivalent and SR 3.3.2.5 Tnp Time Delay SR 33 2.9 7 \
w (a) Vessel AT sg 2,9,n Io
$ Vessel AT
-_k3.1-6)
(Power-1) Equivalent to 13.9% S s RTP) (b) Vessel AT 5 Vessel AT (Power-2) Equivalent to 23.9% i RTP")
..(4) Containment
. 1, 20, 3S 4 O SR 3.3.2.1 5 2.0 psig -
Pressure - SR 3.3.2.5 Emironmental SR 3.3.2.9 Allowance ER 1.1.2.10 l i
. Modifier g 2,9-SC l= A (j) Except when allMFIVs are closed.
(k) With a time delay 5 240 seconds. (1) With a time delay 5130 seconds.
- (q) Feedwater isolation only.
[f
g-ESFAS Instrumentation 3.3.2 Table 3.3.2 1 (page 7 of 10) Engineered Safety Feature Actuation Syster. Instrumentation APPLICABLE MODES OR OmER 3.3 08-SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE Mi@ FUNCTION CONDITIONS CHANNELS COWITIONS REQUIREENTS VALUEW !" "!E
- 6. Auxiliary Fee &ater 3
- p. Nanual]l Initiation IZi3 1/ pump Q $BP 322 NA !
3.3-581-r b; e- Automatic 1.2.3 2 trains G SR 3.3.2.2 m m Actuation Logic SR 3.3.2.4 . and Actuation SR 3.3.2.6 Relays Mohd ' State - !
- p3 . j W ISSED i
- p. b- Automatic 1.2.3 2 trains GR SR 3.3.2.3 m m Actuation Logic :.3.3-59 and Actuation I
. Relays (Behnee i eN BOP p3 l ESFAS)
- e. Sawatantevel Le.w$ew ri) staneaseneratAr M2;p perJG p stuns,2p g
' ;tsesusf 13.3-04 watet!ssvel SRsmI25 NamewjRagge Lwanw SR a ntz:s t&_
EM IF'* _ h $* l Contaiment fff.f.3.10 l r Env1onment) - I2) Stean Sanerator 1.2:;3 4jpecl3G N SRpt323 i31W 3.3 04-Water: Level SR3:32IS . Narrow 8 Range
)
Low ' Low 1(Normal SR3m Instrument Contairment Environment) y,yg,/p Span d f. E I ! (continued) (a) .t . ; c '; ";;;. LW t ;g;i'i; ig'. ...;ti. ;;y _..;i , ;.-ly ',1h .11; t';ig 4 3. .; = 0;t; i .t 3.3-08 St.,dj ;;t M h;y ;;d by it.; ==t The Allowable yelue *finesithe:ltaiting safety?systeersetting.T*See the Bases for the Trip Setpoints. 84 j ( L HARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 43 5/15/97
-r .- _ _
o l ESFAS Instrumentation 3.3.2 l - Table 3.3.2 1 (page 8 of 10) Engineered Safety Feature Actuation System Instrumentation APPLICABE HDDES OR omER 3.3 08 SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE RIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE" TBPGIW"
- 6. Auxiliary Feedwater (continued)
LU Vessel-ar 1.- 2 4 M sRs:3 2n 3.3-04. Egtrivalentsend SRI 3:3'2'5 ltjp*I1te1 Delay 5nx CJ L
~ ' ^ ^
ggg,g,2.jp R12M Eqir -'--- ' r s vesse pT S 23-27 r=2w. Matvslent 7.y,,7 E siiipBRINP-- ton 3M 1 15E m essel W HP")
!PMPomer30 j N M***#W CA-27-dr7 E"EwesseMEiF-- - Eip1 valent isr _'"- r' " to:Es s 9,7-/g
!B iiiiiAB88485--- BIP*
.. 2"3HesseltllLT l C2Xtower3)
(4) contatnment n;;:r;;3 a p sitss;231 k;2;ogags ;3.3 Pressu rg SIG5225
= " ****o - -
Ewitenments] - gewarri~ _ gyf,Q.l0 $ SS WU llodZA g -- esd- Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements. 7:e- Loss of Offsite 1.2.3 2;tnains F SR 3.3.2.7 t-29tM t-295-V Power iF per SR 3.3.2.9 .;itt. ; 0.0 wit' :.: B-PS bus = 0.0.2. : see44 e see-Hee 3.3 55 de+ey NA *4ey
- f. ..:_ scit;;; .;ter - he e-per + =.3.2.' . =t L: 2 70; L:
= =t r g bes = :. .2.; veMese vehese 3.3 01
= :. .2.10 (continued)
(a) = si=r : =tc . Unit .c.ific igi_....ti=; ny _...:i =iy "1uti; ";1.; t.e..::r.; an RtFM 3.3-08
- .dy .att.,t~ h;y ;;d by tt; nit The AllowablCValueldefines3he Maiting;safetypystem; setting.'J:See the' Bases-for 'the'Tr.ip'Setpoints.
Ik) Withia' time delay E 240 h nds' . l (1) With a time delay .s330;secorxis- 3.3-04 MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.3 44 5/15/97
I ESFAS 2nStrumentation 3.3.2 i t . _ . . 1 Table 3.3.2 1 (page 9 of 10) i Engineered Safety Feature Actuation Systee Instrumentation l
.i APPLICABLE HDDES OR [
omER 3.3 08- l SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE MW. ! FUNCTION COWITIONS CHANNELS COWITIONS REQUIREENTS VALUE" GRp01W" !
- 6. Auxiliary Feessater ;
-(continued)
- g. Trip of all Nain 1.2* 2 per .1 SR 3.3.2.8 2 - pe+g 2- -pe+g 3,3 16; Feeerster Pimps pump Oj }.{.{.{ NA
- 3.3 55 -
l
- h. Auxiliary Feedwater 1.2.3 23 FF SR 3.3.2.1 amt!64Dels ; ;;;;
- .3.3-17z l Pimp Suction Transfer O 2.3.2.' i 20.';; ;;i;
- on Suction Pressure - SR 3.3.2.9 Low sugs2;22 "3 3-14"
- 7. Automatic Switchover to Containment Susp
- a. Automatic Actuation 1.2.3.4 2 trains C SR 3.3.2.2 NA NA Logic and Actuation SR 3.3.2.4 Relsys [SSPS) p3 l
- sg gr.2,13 O f.1-M i
- b. Refueling Water Storage 1.2.3.4 4 K SR 3.3.2.1 km H 2---and < t. g .ps . ?
Tank (RWST) Level - Low _ _ SR 3.3.2.5 =d . t . Low SR 3.3.z.9 (.ch+; renn .ch,%- c =.=.w i Coincident with Safety Refer to Function 1 (Safety Injection) for all initiation ! Injection functions and requirements. ! 1 (continued) (a) % .i;.;r'; ";tc. .'t h;ifi; i;;l-...^..ti;; ny -..;i sly ',l ti; '.';1.2 . d r., ; 0,.'., ; t
. ,3,3 08< !
St.dj ^...J ,leif .x d by tr.; ..'t ThaATjenable Value'idefines3he31 siting;safletymystemtsett%g@ thef. Bases;for$TdgWints _ _ i
',J m _ itia 2 (n) Trip # unction aey1mltiocked$t' befort shutdown!ofD3ast operating esiniesdwater-pump 4ndh 3.3-16 just !after*the 'firstrasin'fecesster; pump :Is. ,put dnto jservice dollowing;perforsance.cffits Wrtup_;tdp test.
Q 3.1-I9 MARKUP OF WOG STS REV 1 (NUREG 1431) 3.3 45 5/15/97
1 l l LOP DG Start Instrumentation 3.3.5 , SURVEILLANCE REQUIREMENTS j SURVEILLANCE FREQUENCY SR-3.3.5.1 P;bfere C"/?S::L C: CK. De]tmeatensin@weenJ80 a 12 h;;r; 3.3 28 ; Vac[Mses110001?and'31G03?andMRac fjgleys busesjjiG021and M 041shaTi h Werjf43djegej j I 1 SR 3.3.5.2 Perform TAD 0T. Eldays B- I i SR 3.3.5.3 Perform CHANNEL CALIBRATION with setpew* [B months B PS Aliadl; '.'el;; [gipM,M
.Nige as follows:
RI ULMS M l BlNEiliMUNBV3RillBEREBBR! ster
!!CIB2Biluit:25 tense ux Y. l4'k*2
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1 Containment Purge and Rh;ust Zsolation Instrumentation PS 3.3.6 OmininmenO fu . ..=, < . t ryl .frm E s p 5 e &S.3-SCh SURVEILLANCE REQUIREMENTS # EIN A/I E N E
.................. .................N0TE.. .....-.... .......... ...............
Refer to Table .3.61todeterminewhichSRs30ther2 thy, . 3.3-55 geMrntion: pply for each Containment Purge and Shaust Isolation Function. PS-m : rngegifiLtatTon't; _ . 'SR75?3Q SURVEILLANCE FRE0VENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours
- A/rrE -
-rhe conh*ner cheeky k exc/ud'ao(. 2.3-39 SR 3.3.6.2 Perform ACTUA ON LOGIC TEST. 31 days on a MII'#
STAGGERED TEST l BASIS 3T-39 6V.Jzytr fepof9Astia4AYfESf2) J $23-7:2-1 s-rq
- ~3 Q3,3-3:3.
SR 3.3.6. Perform COT. 92 days Vf/fff? ,= xs-r9
$ 2 7-3'l.
n M n .
)
[ ,- . e; sh (continued) , MARKUP 0F WOG STS REV 1 (NUREG-1431) 3.3 62 5/15/97
l Containment Purge ar.d 06aust Isolation Instrumentation PSc j 3.3.6 . l l l . . - l - SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY SR 3.3.69 -- -- - ---
--- NOTE-- - - -- -- -
[ Verification of setpoint is not required. u.h ..................................:............
- g. e-n Q 1,2-12 Perform TADOT. IB months B-2.2-19 S & 2.2 SR 3.3.6T P~erform CHANNEL CALIBRATION. IB months B.
g_ __ , f k'$$21 m - - - - - --- . ye ;iaii.iaGin Y?NW GirlWtian EM
~
lBgesmesesisi :3.3-31r l
.?.*F-SS x : \ I"
(. 4 I a MARKUP 0F WOG STS REV 1 (NUREG-1431) 3.3-63 5/15/97 I i
l l RTS Instrumentation h , ., B 3.3.1
~
l l l BASES i l l ACTIONS YTandT25Jcontinued) l Complet.jongJme3fj61 hod.ps2fsMsed30plReference7g5Ef3;he Inop~e nable;channeEcannot31xqplaced;gnihf3tgleengtMnattj!hth th!0specifieAdmpletionmmetthe:tuatimusube3gustga!ajag wheneJthisyynetiorgsstmeniM!dDMmiLusissan pdditionaysghourswallowedsp]acepanst2rt'elmEE 1 Slf,tVEILLANCE The SRs for each RTS Function are identified by the SRs column of - REQUIREMENTS Table 3.3.11 for that FunctionM $f,1-SF A Note has been added t; the SR T;ble stating that Table 3.31 1_ - 6 determines which SRs apply to which RTS $3.7-SS Funct1ons. Note that each channel of process protection supplies both trains of the RTS. When testing Channel I. Train A and Train B must be examined. Siinilarly, Train A and Train B must be examined-when ' testing Channel II. Channel III, and Channel IV. (if ;.ppliable). Q*Q* The CHANNEL CALIBRATION and COTS are performed in a manner that-is consistent with the assumptions used in analytically r.alculating the required channel accuracies. Revic.;;r * - A t;; Cert;ir Tr;quxic:; er; b;xd = es;nd topicel . g ert;. Ir. erder for ; lic;r.x; t; ex trax ti;;;;;. th; linr.x; ;;;u:;t jntify the Tr;;nicic; x r; quired 53 tre :;t;ff SER fer the topic;l .;p it. SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an ! indication of excessive instrument drift in one of the channels l or of something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying that the l instrumentation continues to operate properly between each N CHANNEL CALIBRATION. . l t, (continued) MARK UP 0F NUREG 1431 BASES B 3.3 60 5/15/97 I l
RTS Instrumentation B 3.3.1 l
- s....
, 'y
. . BASES SURVEILLANCE SR 3.3.1.14 (continued)
REQUIREMENTS The SR is modified by a Note that excludes verification of i
- setpoints from the TADOT. The Functions affected have no j' setpoints associated with them.
exceell -/fe f in.js,.fock SR 3.3.1.16 8"' Y'
'^ MQE3. 'Tku."II'
"# W i SR 3.3.1.15 is the performance c f a TAD 0T of Turbine Trip Functions. This TADOT i; x dc cri k d in SR 3.3.1.4. ; n;pt tk t this t a t is performed prior to t M N
@ Surveillance is not required if it has been performed within the previous 31 days., Verification of the Trip Setpoint does not have to be performed for this Surveillance. Performance of this test will ensure that the turbine' trip Function is OPERABLE prior t g .
act er cal in t c no a a and st o l a st. up exceedi 'l$a A-f In1htrbek, y SR 3.3.1.16 _ G .3.fS5~
'iR 3.3.1.16 verifies that theGfefufWJindividual channel
;ctuation response times are less t1an or equal to the maximus -
ve. lues assumed in the accident analysis. Response time test 4ng ggi
.p.r r.c.c.;t 6 acceptance criteria are included in Ta! .isi "equir; r.t; " anni . Irm A ndividua I
-~
component response times are not modeled in the a lyses. . The analyses model the overall or total elapsed time, rom the point ' at which the parameter exceeds the trip setpoint alue at the ! sensor to tk p; int et which th; equi, ent in;.y th; r;;; ired I f;retieni ;t;t; (i.;.. .entrol ad ;btde , 73 fully inxited
..iJ1.1 in-th; rs;ter ;ere). Ingg@oss".cf2stathemgig#
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Yahan 44 e*rmse inhekmts /Mh4E.$<.rme Mr /ae N. .o , For channels that muuur uynamn. tranuer own uuns 7.g., lag - i lead / lag rate / lag, etc.), the response time tat syg k verification'is performed with the tra;f;r F;retien j.at te-ene withtkraultingnn;r;drapaxtin;@redyth;
;ppr;priet; FS'" resp;nn tin. Alterr.;tcly, t k pnx tin test ;;n k Frierr;d with the tin anst=ts n' g; their .
n;;i ni vel;;s. pr;vided th; r;quir;d respenn ^y; is .
- l. _~ 1 s (continued) pgg ,
MARK-UP OF NUREG.1431 BASES B 3.3 69 A/, sw Ms 5/15/97 R& ment.r
.
- fv 4aee W.A. ir
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. .- - - .- - - ~ - , - .- . _ - . - - -.. .- -
i RTS Instrumentation l B 3.3.1- ' BASES l SURVEILLANCE SR 3.3.1.16 (continued) ;
- REQUIREMENTS
- n: lytic;11y ;;lculeted ;;;u;;;ing the time constants are set at l
their nominal values. The response time may be measured gggget! . by a series of overlapping tests;2prjot_h'erlAgdgRtiatinhWigg { Refg'9)J such that the entire response time is RRW,plad! l
;;;;ur d. I flespgIlse,jtin@gMvergiedAVjAcMLa]m nyasetgessfgsamentlamioreglappingpfttstegenias '
. - - 1,~ n .
MMDit!gacgargsportsegene'4esigantm!ggianattSGERLe chann reWLocationsifar,wnsorminessmaGlassig53RBlanggi Dinal&Dat!@MeJmes.pngsLangagguaggggggummage teststytogggggonen-On8Repr@H85ENele2kinagow GO.LitRtz,snitigggem.te,_ggaenmm.. em . . i.am w.mneem m_ m .. . ..__.e ta - n= ...mm----. gm.... ..,
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talaemome_ m tviey - w 3 esy z : s -s . j DePlMiggEhepens!gggssembbmJpf M Inyaditgrga11odst46nsNor ottheriantt3tomaggs!!amistan_s ofathe3agitteppenseteL*!Ly_1bewse!U!Lreengututlis measuigi.i.itasing"gheyethodologyppfROJ)PASES2MBBitgBigggggl abote7dnDecordanceggMj293MettagglBE As appropriate, each@% channel's response tjue must bi 3'3-SS verified every 18 months on a STAGGERED TEST BASIS. E4 rih per.1fi cationj~shal19 ncl.edelatp ea_st;4ne]tRigisGt3Qtg8Mt trainsgarey erf g edMt3 east /pncej er M M ,2 Testing of the final actuation devices is included in the t;; ting. gejiticatllag? Response times cannot be determined during unit operation because l@ l ! (continued) i MARK UP 0F NUREG 1431 BASES B 3.3 70 5/15/97 q , -v ,-
. -. - .- - . . - . . - . - ..~ .- - - - . - . -
ESFAS Instrumentation l B 3.3.2 BASES ACTIONS P.T'and ~P:2;ilcontipued) fot?thetAFWipumps EHhisl$functionihasfa3 %tytKtrApllgggg ) ThegefoneMJitiquedfopeption::Isyallowed:with'ionegacyggljie l phagneBuntilithe lperfonsagoeyotigheinent#sonth]y Miofggg3g thelbtherichanneMias3 ong As @helinopgabjeyhanpdM1W
$21pMthinI13houmConditiongM s3modifjietilbJaiMidtpMat33 thatILC. 0,3@Xis3mt[applicableAgDDEJyh_aggggenpyelpgggi ;
kithjagJ.1gop,egab]G:hanne)). j Qa pandigonRamy_eszpphe%ursuawm ! Eunctstwnstth3agegnonp3*ast11st3magnagggaeurasan RPpUtableMienditjientsMandjBegial.gedEgig!lg '.WO.9 J CA-3/JNw2 ga g essoeptedstos u a y l eedwater w i m .e m
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4~,.% RTiandit2 Dj- ' pendittorL3ppjgs39]s: hem 9WRitLidgatagle ,9 e .. . ESfFAMc3@gMgggg1.cRrtdaith tEajn3nogegetite,N_3et.;; @;a ,0 at.s ,,: DioenswM33EtttkittieMtiwl A 2ilNo o
$ggjgqRge_ M ',Jyr . ;-
M M 3 11g19800]E@r2Rg3Legt s M w 'c . I!PJinitsig3tegggsm SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs column REQUIREMENTS of Table 3.3.21 6 5 % Q?To ff,7-56 A Note has been added tc the S", Tcble to clarify that Table 3.3.21 @ determines which SRs apply to whic ps.7M ESFAS Functions. Note that each channel of process protection supplies both trains of the ESFAS. When testing channel I, train A and train B must be examined. Similarly, train A and train B must be examined when testing channel II, channel III, and channel-IV. tif c;pikebic). The CHANNEL CALIBRATION and COTS are performed in a
-Q (continued)
MARK UP 0F NUREG 1431 BASES B 3.3 138 5/15/97 Tr
l ESFAS Instrumentation
, B 3.3.2 SURVEILLANCE SR 3.3.2.9 REQUIREMENTS (continued) SR 3.3.2.9 is the performance of a CHANNEL CALIBRATION.
A CHANNEL CALIBRATION is performed every 18 months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology. The differina betuxn the current ";s f;und" alua en the previous test ";s left" niacs rast be censist;nt with th; drift ell;;;na u;cd in th; xtp; int n thedele;y. The Frequency of 58 months is based on the essaption of en 18 enth assumed calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology. This SR is modified by a Note stating that this test should include verification that the time constants are adjusted to the prescribed values where applicable. B@s%etsmotMncidae vergicationJMme *W3elatsGhese3nreDe11fierma responseRime9estugM23:430] rot e r- ra 2.2.a.9 x M.H+ SR 3.3.2.10 e u _ig._ Q gg, Gen ] This SR verifies the indi idual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis. Response time verification M __
@ acceptance critoria are included in the Techni=1 p y,y_SS-R; quire x at; .";n=l"(Ref. & Individual component respons times are not modeled in the an yses. The analyses model the overall or total elapsed time, om the point at which the parameter exceeds the trip set: int value at the sensor, to the point at which the equipment 'l both trains reaches the required functional state (e.g._, pump at rated discharge pressure, valves in full open or closed posi on).
For channels that include ynamic transfer functions (e.g., lag, lead / lag, rate / lag, etc.), the response time verification $s may
*+ aulyres y Q ;HL y e & cfr w7e'gtn,g THAIo credif w MARK UP 0F NUREG 1431 BASES B 3.3-143 , y g5,/1 -
_ YArdOW78
f-CHANGE NUPEER JUSTIFICATION Given the above changes, all instruments monitoring RG 1.97 Type A variables are included as well as those Category 1, non-Type A variables that meet Criterion 4 in 10 CFR 50.36(c)(2)(ii). 3.3 22 Consistent with current TS [3.3.3.5], hiseDdHD h',Jft9H2WarrT,hr.manQreactor trip breaker and RCP breakerp g7 22 . indications are excluded from] CHANNEL CALIBRATIONS in ITS SR 3.3.4.3 since these SRs have no meaning for [these] functions. 3.3 23 Not applicable to Callaway. See Conversion Comparison Table r 3.3 z4 lirdedIn.rfrums fadron at P nchans and ' This change vises improved TS Table 3.3.41 to reflect current TS [$r.v22f] deletes references to " controls" in the tablemr s'r Q 27-J+
= 3,3 fg y ,e nn.e ag w u.e e Enti;7,;- ;; 7:g;;
P =' *" ri .:: th tdic t L;.. . .. ;;;d te 'n? d ;r,13
- Mt--- .t: tis. As r ai;ed. tra t;L1; i; r.e m,. .;t:..t gitt, tr.;-
l
": -t ;f tra W, . teL1;; ir, I"; Sectis 3.3 ...d i teJ .!cil, ;t '
n it.; c., ...; T'; . 6 7; ..;:. . r M .r E A* 7 - [, M 3.3 25 This change adds an Applicability Note consistent with that found in improved TS Table 3.3.11 (i.e., source range neutron flux is only required below the P 6 interlock). This is consistent with current TS Table 3.3 9. 3.3 26 The Pressurizer PORV and Block Valve Controls are deleted and RCP ; Breaker Position and AFW Suction Pressure are added to improved TS ' Table 3.3.41, consistent with the current licensing basis for compliance with GDC 19. The PORVs may be used for eventual plant cooldown: however, they are not required to attain HOT STANDBY which is the basis for the listed functions. The added functions may be used to ensure decay heat removal by the SGs in attaining HOT STANDBY. 3.3 27 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).
.ruryeillanc e. 4 T-2 3.3 28 V Tie breaker 6 per current T5_TableDdLKjeU0r3tetJoefit49)for Functional Unit 8.b added as @fots),e95^1X1Le6F&thr1LanMO SS,7.28' dDepttTank:3)ITS SR .3.5.1.
/.r 3.3 29 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).
.1mr 3.2. 4 -
3.3 30 The portion ofVCondition CIfr61ahnfaVa##Mreferring to Q 3.7-70 one i or more rable functions is revis with or to cover more automatic actuation trains one, BOP ESFAS a e f ma M n "f h Tn [ onl r To e y
- 3. 31 u requ re s 5 ef
[undervoltage and degradea voltage diesel generator start functions and t containee rge isolation functions to be verified against theg gs 55 JUSTI THM " O!FET" CTS'~ TS
- 5/15/97
CHANGE NUMBER JUSTIFICATION l . l 3.3 49 ITS SR 3.3.1.8 is revised to extend the conditional COT frequency for l power and intermediate range channels from 4 hours after reducing power below P 10 to 12 hours, based on operating experience regarding the time needed to perform the COTS. It stands to reason that if 4 hours are allowed for 2 source range COTS 12 hours should be allowed for 6 intermediate range and power ra69e COTS. The SR continues to assure that the COTS are performed in a timely manner after the requisite l plant conditions are entered. This change is consistent with traveler i ICC. T.TTF' ;2+.;2 Q3'S M W ?.L": ; 3.3 50 ITS SR 3.3.1.12 is deleted per the current TS. Where cited in i Table 3.3.1 1. a change to SR 3.3.1.10 has been made. The plant- ) specific design does not use bypass loop RTDs for the protection system. 3.3 51 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68). 3.3 52 This change adds Note [(1)] to ITS Table 3.3.11 per the current TS as an operator aid to note the dual RTS/ESFAS functions of SG Water Level Low Low. This is an administrative enhancement to reflect the plant's dual requirements for this function. 3.3 53 Not applicable to Calla y. See Conversigp Compgrison Table Q, (Enclosure 6B). *fa n/d[M 2.TM,1*~l delehe N 224 // and #274 /3 v - 3.3 54 Function 18.b (P 7) of ITS Table 3.3.1-1 is clarifi COTS and bTc OneS'an T Channel Calibrations apply to the P 10 and P 13 i s not to the P 7 g/ gj j,, logic function.+ This change is an administrative clarification to 43.7-S/- g ggfgy* address he relationships bet these interlocks in the' ant's , Cr:r.rueve r//ance r-f-7 i dysf
- er.e ' no corre.rf
[fy, ,,, y,.ff 22/- 1*Id Mon 3.3 55 one umuve ievises us arna.a.2.io anuga a.a.e.lu To verity
-% sponse timeg. accommodating those channels ave no S 7,*J. S C i espnx .. _ mmiru.us per the current lic basis. [As such, g!
g4 line item referen o these SRs in .3.1 1 and 3.3.21 can be g._7 j deleted. The current TS ntify which functions have a e /r, required response ti he res e time limits and the ggg7.g identificatio hose functions that required response times are curr isted in@ licensee controlled doc not in the current
. This change makes the ITS consistent with this a of the current TS. similar revision to the ITS 3.3.6 SR Note h o been maderegardigSR3.3.6[.] 2 -D 3.3 56 This change revises 115 3.3.2 condition J to reflect current TS l Table [3.3 3] Action Statement [19] for Functional Unit [6.g]. [The l g required ACTION is to place the channel in trip within 1 hour rather j than to restore Operability within 48 hours.]
3.3 57 Not ttsed - /re,j/c -/s O //,v4 , $ gg y7 y 7,3 yg-m JUSTIFICATION FOR DI ERENCES we u).TS 7 5/15/97
l-INSERT 6A-7 Q 3.3-55 l l This change revises ITS Tables 3.3.1-1 and 3.3.2-1 to delete response time testing surveillance listings (SR 3.3.1.16 and SR 3.3.2.10) for those Functions that have no response time testing requirements per the current licensing basis. The response time limits and the identification of those Functions that have response time testing requirements are currently l' listed in (a licensee controlled document. A revision to the ITS 3.3.6 SR Note has also been made to accommodate SR 3.3.6.6]. l 4
kt h, ( j
~
CONVERSION COMPARISON TABLE FOR DIFFERENCES FRON NUREG 1431, SECTION 3.3 Page 6 of 21-
. TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON C(MANCE PEAK WOLF CREEK' CALLAWAY
' ~
3.3 31 The current TS reasire the response times associated with Yes- ~Yes . No Yes
% [undervoltage and aded voitage Wse generator ^ s l start functions and t containment isolation d 2 h function @o be verified against the specific response y $ 7. 7- 5 6 une valueTi.unismeo in tne Lt:At]. The ITS is revised / ,y ,
to match the current TS and the response time values are [ retained in the FSAR]. As is done with the Reactor Trip l' System and the ESFAS instrumentation, this method is an appropriate way to control response times. [SR 3.3.5.4 andSR3.3.60] are added to require the response time verifications. 6 [, ggg 3 j 3.3-32 Isproved TS @ anie 3.3.6-1 is changed to be consistent with current TS [3.3.2 Functional Unit 3.c and current Yes [] No - see CN 3.3-73. Yes Yes TS 3.9.9]. Subfunctions [b d] of Containment Radiation are stricken since only the gaseousg channels provide k 2.'I~I2-the actuation function [and the bracketed setpolet IS changed to reflect plant specific requirements]. i ogi as n see ne c it - AS t s I ist SR 1pdmen in 3.3-33 Improved TS Table 3.3.7-1 is revised to reflect the plant No see CN No - see CN 3.3-78. Yes Yes design. The CREVS is actuated by radiation monitors 3.3-102. located in the air intakes, by a containment isolation -
! Phase A signal, by a containment purge isolation signal,
, by a fuel butiding ventilation isolation signal, or manually. The bracketed setpoint is revised to reflect current TS Table 3.3 6 requirements. 3.3-34 Revisions to add an LC0 3.0.3 exception Note 1 to Yes No - LC0 does not Yes Yes ITS 3.3.8 reflect cawyed7T apply. $ 7,7-Sf-3.3 35 Improved TS Table 3.3.8-1 is revised to reflect the plant No - not consistent No - LCO does not Yes Yrs design. Only the gaseous channels provide the actuation with plant design, apply. function. The bracketed setpoint is revised to reflect : ) current TS Table 3.3-6 requirements. l CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97
1
~
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 SECTION 3.3 Page 9 of 21 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.3-49 ITS SR 3.3.1.8 is revised to extend the conditional COT Yes Yes Yes Yes frequency for power and intermediate range channels from 4 hours after reducing power below P-10 to 12 hours, based on operating experience regarding the time needed to perform the COTS. It stands to reason that if 4 hours are allowed for 2 source range COTS, 12 hours should be allowed for 6 intermediate range and power range COTS. 3.3-50 ITS SR 3.3.1.12 is deleted per the current TS. Where No - see CN Yes Yes Yes cited in Table 3.3.2-1, a change to SR 3.3.1.10 has been 3.3-101. made. 3.3 51 ITS ACTION B.2 of LCO 3.3.7 is deleted, since DCPP cannot Yes No No No cperate CRYS with both pressurization systems running at the same time. 3.3 52 Added Note [(1)] to ITS Table 3.3.1 1 per the current TS No - adopted ISTS Yes Yes Yes as an operator aid to note the dual RTS/ESFAS functions of fonnat. SG Water Level Low-Low. 3.3-53 The REQUIRED CHANNELS description for Functions 2.a and Yes No No No 3.b.(1). of ITS Table 3.3.21. are revised per the DCPP CTS to note that only two switches (one per train) exist and that both must be moved coincident for manual initiation. ( ~/*
- dd JX 3,3.U"an efe h SA S3/.// edff.xgf, fy, 3.3 54 Function 18.b (P-7) of ITS Table 3.3.11 is clariri Yes ~hu Yes COTS and Channel Calibrations apply to the P-10 and -13 -*es.-% avvpteu m r inputs, not to the P-7 logic function.
fm .m .- Vq 6 .~f.3--SV-
/
3.3 55 '".gg' Rev res 3.3.1.26 and time ac 3.3.2
^
to verir ating tho che s
"; T: ei" h ide.cd in ii3 -
Yes Yes Z: Will 6 Yes ret:i=d in IT"r
'that ye no r .ponse r rements per he cu ent Te k: f;r rcw; red S 3,"S 11 sing b s. [A suc . ne-item re rence o the Tales T.~uvo;. ivr .[sa; red - -rw ~ u un. .
s in Ta es 3.3. -1 and .3.2-1 can dele A { I
~
ega i 3. .h] ar- (mi-fr e A ' SERT' 48- G 3.3-56 Revise ITS 3.3.2 Co t1 J to reflect current TS Table No - See CN Yes Yes Yes [3.3-3]. ACTION Statt [19] f3r Functional Unit [6.g]. 3.3 116.
' hI CONVERSION COMPARISON TABLE 1431
?
5/15/97
, INSERT 6B-9 Q 3.3-55 i This change revises ITS Tables 3.3.1-1 and 3.3.2-1 to delete response time testing i surveillance listings (SR 3.3.1.16 and SR 3.3.2.10) for those Functions that have no response time testing requirements per the current licensing basis. i i I i e l f, h 4 i I t i i l I l i l' i l 1
. ~ - _.__y. _ . - _ _ _ _ . . _ - _ . _ - _ _ - .
l i ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q 3.3-63 APPLICABILITY: CA, CP, DC, WC
]
REQUEST: Revise ITS Table 3.3.2-1 [ Note (g)) per current TS Table [3.3-2] tJote [b). This revision is a clarification to the operator that describes the circumstances under which the [Steamline Pressure Negative Rate-High function may be blocked when below the] P-11 permissive. Comment: The wording of the applicability notes (b) and (g) in the ITS is more limiting
- than the CTS applicability, therefore, the change is more restrictive and should be documented as such. The CTS notes that the functions must be operable in Mode 3 but may be blocked below the P-11 setpoint. The iTS does not require operability below the P-11 setpoint in Mode 3. The blocking instruction to the operators belongs in the Bases.
FLOG RESPONSE (original): NRC deleted the reference to note (g) at the 9/15/98 leeting. ITS Note (b) has been revised to reflect the intent of CTS # note requirements for D. ? Callaway, and WCGS, and the (a) note requirements for CPSES, as described in new DOC 2-22-A. ITS note (b) now states that the Functions are required above P-11 and are required below P-11 unless blocked. ITS Note (g) is revised to reflect the CTS N note requirements for - DCPP, Callaway, and WCGS, and (b) note requirements for CPSES. ITS Note (g) now states that the Function is required below P-11; however, it is blocked below P-11 when Si on Steam Line Pressure-Lcwis not blocked. In addition to these revisions, the CTS M note for DCPP, Callaway, and WCGS, and (b) note 3 for CPSES. are revised to move the descriptive information related to the automatic blocking of ! these functions to the bases via DOC 2-50-LG. FLOG RESPONSE (supplement): Based on a telecon with NRC staff on January 12,1999,
, ITS Table 3.3.2-1 note (b) is revised to read: "Above the P-11 (Presur!zer Pressure) interlock and below P-11 unless the Function is blocked." No additional Enclosure 2. 3A, or 3B changes are needed from those submitted in the original response. The Bases discussions for the affected Functions have also been revised to identify which signal is being blocked.
- ATTACHED PAGES:
Attachment 9, CTS 3/4.3 - ITS 3.3 i
- Enclosure 5A, pages 3.3-37 and 3.3-40 !
l Enclosure SB, pages B 3.3-87, B 3.3-89, and B 3.3-103 j l 4 h
ESFAS Instrumentation 3.3.2 Table 3.3.21 (page 1 of 10) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER 3.3-08' SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE MiP FUNvTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE" SEPGWT*
- 1. Safety Injection
- a. Manual Initiation 1.2.3.4 2 B SR 3.3.2.8 NA E
- b. Automatic Actuation 1.2.3.4 2 trains C SR 3.3.2.2 NA E Logic and Actuation SR 3.3.2.4 PS l Relays USPS) SR 3.3.2. 1 st s.s.a.is & 3.3-M
- c. Containment 1.2.3 3 D SR 3.3.2.1 kM.5Dsis . 3.0 pis B PS Pressure - High 1 SR 3.3.2.5 6-66 .
SR 3.3.2.9
; ; .J . 2.1"3 ;
1
- d. Pressurizer 1. 2.3
- 3p D SR 3.3.2.1 E t834 M i 1:0 pig g,p3 j Pressure - Low SR 3.3.2.5 1839 ;
& 2,*3- % h eL Y rC P1 rf e -
V* evW
- e. Steam Line Pressure "MiD +1+ Low 1.2.3" 3 per D SR 3.3.2.1 !ESE12istf e-676 B PS g steam SR 3.3.2.5 635 ps4g* l line SR 3.3.2.9
. 0.2.1 D feHHgh hEr3 B-per 0 = . .2.1 s 100 pig . = pig 3.3 01
- fec;c.tici etene = :.3.2.5 pressure we = . .0.3 )
0;t;xn :::a = :.:. le l t4fes j 1
; "ig sic;; ils.; " hea
- e-per e = . .r.1 fe+ ff+ 3.3 01-Two-St;;; L1=; stene 0" 3.2.2.0 line 0",0.0.0.9
.3.2.10 i
C;i ai i..; .:;th haa* i-per 0 = . . .1 . 20.5 r . ::P r i T ,- tow-tow 4eop = .2.0.0 3.3 01 l 0" 2.2.2.0 l
= 2.2.2.10
[(continued) ; l (a) Ocv;;;;r'; Otc. Un t . w ;f;; ;;pi..~ .;et;;n; ;;y a .;;;. enly '.il;.; M ; "; lac e g ad;ng en 0;tp; rt 3.3-08 St d., ;;t b il;;j x d by t M an't The'AllowableWal Mefines'the tipgts fety3ystes:settingF25ee the sases ;for'the' Trip 'Setpoints; Me neffen /r g,g (b) AbovetheP11(pus'surizerPressure) inter ek[anM l>8Iov f-ll gnle.rd/sedc/, ,743 g (c) Time constants used in the lead / lag control .r are t 2 50 seconds ard i, s 5 seconds. (d) ?bere-tk " : (T,.,- t;.; L;;; interi;;k. - A ful' st; a fle kle.; 200 1;;d. c.nd or ic; Less tM.; ;r ews.1 to e feation ifimd es or ;vci;.,p.4.ng t; 't 3.3-01 inereesing limerly fra '4: full ;tca 'le.; et 2001;;d t; 1140 ful' ;tze; ib.; et 1000 leod. end o" l ;;; . ;;p.. ding t; 1100 ful' ;tca 'ic.; ek.; 100: hr.d . l _.,,., (f; Less tMa or twel t; e function ifimd ;.3 o" wiriap;.d;ng t; 400 full stce; fic.; bct un 0: and 200 lead 3.3-01 l w- ____ enHMn ; or = a; eain; v; '- ' ;; 30: st;;; 'b et 20: bed ;; 1100 ful' ;te;; ;.; ;t 100: h ad. fl I. . k '.1K62 .?K630 .K740 and 741r . ch. s 1U ltes yev JJ18 ns; no pr)or A,K19/ j erj 'M00
- er) .anit~ s lin - j5 r;6; r.y 4 hou . ?1 t" o wit)fn- f f l
0 3.341 MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.3 37 5/15/97 l l l l l
l ESFAS instrumentation 3.3.2 ! Tabic 3.3.2 1 (page 4 of 10) Engineered Safety Feature Actuation System Instrumentation APPLICABLE HODES OR OTHER
-3.3 08 SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE MIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE" EETPGET"'
- 4. Steam Line Isolation (continued) l
- b. Automatic Actuation 1. 2"' , 3 "' 2 trains G SR 3.3.2.2 M M , p3 ;
Logic and Actuation SR 3.3.2.4 j Relays [SSPS) SR 3.3.2.6 , 1 r.: Automatic:Actuatson m2"*;35 g 9 .sa 4 ch e:s r - m M gggg 3.3-98 Logicannd Actuatson trains
- 2 F.I. 2,2 Reles3(ftsgSJ j
p; e- Containment 1,2") . 3"' 43 D SR 3.3.2.1 i!:;18;3.jpsig r-6-35 g.p3 Pressure-High 2 SR 6-61 peig (.,n 0.0.2.;D
& r.s-ss- Ac+;ec~e 4 pfsp, n d- Steam Line M[a- /[,,u ;# y 1;. Pressure ]
(1) Low 1.2 "> . 3 per D SR 3.3.2.1 k$llW8 r 95 B-PS' line 2. C T 3.3.0-1 (2) Negative 3 *'"' 3 per D SR 3.3.2.1 )RilR4*gpsj x-iH g.p3- I Rate - High steam SR 3.3.2.5 Hb6 x pr# see* line SR 3.3.2.9 psWeee
@ .2.2.1")
- . "i;'. St::: n; i- 5; h2"',-3"' e-per D Z 0.0.0.1 fe) fe 3.3 01 St; ; Linc; eteen 2 0.0.2.0 44ne Z 0.0.2.0 0 0.0.2.10 0;in;i;,...t .;ith h?"'- 4-per D Z 0.0.2.1 2 ZO.0T r-SSPF T,.,- Low-Lew 3"'"' loop
((.{.{
*' W %...
(a)
.m..m.~ . .. m._..m..
. . _m._,._--m2_f....,...._z._.
..z.. , . . . . ..
-.,,. -.T.....
._< , _ ,,-r.
_ , . _ m -~~I_,$.
,~ -
. , _ .' 3.3-08
" "* ' ""d '
Study ;ctt.;;;hi, u;;d tj th; unit The%11 able V61ue; defines'thejimitingtsa . settingdee the' Bases; Tor,.%eHrfg$etpoints. ~3
#[e/en,h//un/e.t.r# [p Me N S.7-43 (b) Above the P.11 (Pressurizer Pressure) Inte le ' . g ?.f-43 (c) Time constants used in the lead / lag controlu -rsec6nos and 1, s p s R w e -
B (g) Below the P.11 (Pressurizer Pressure) Intericc, 9ayM;blpcked belowf31Nhen4afetyjinjection 3.3 63-onJow' steam Qine pressure;1s;not", blocked.
. . , (h) Time constant utilized in the rate / leg controller is -- ) seconds. 3*3 12 (i) Except when all MSIVs are closed ;nd dc ;;t".eted- B PS.
(0) Each train yequires ,a minimum of two.Trogrammablejogic; controllers to.*be't)PERABE; 3.3 98 MARKUP OF WOG STS REV 1 (NUREG-1431) 3.3 40 5/15/97
L-ESFAS Instrumentation M, B 3.3.2- \ g-j m l- . APPLICABLE d. Safety Iniection Pressurizer Pressure - Low SAFETY ANALYSES, (continued) , LCO, and l APPLICABILITY
- A. spectrum of rod cluster control assembly ejection accidents (rod ejection):
. Inadvertent opening of a pressurizer relief i or safety. valve: )
. LOCAs: and .
1
. SG Tube Rupture'3
. , . l ::;. a .. . .;
--^
T pressurizerpressureprovidesbothcontroland@836'^~I protection functions: input to the Pressurizer Pressure Control System, reactor trip, and SI. s Therefore, the actuation logic must be able to (re withstand both an input failure to control system, which may then require the protection function actuation, and a single failure in the other , channels providing the protection function i
' actuation. Thus, four OPERABLE channels are I required to satisfy the requirements with a i two out-of four logic. Ter mits th;t h;n dediated.pretati; ed catrol cher.7;l;. =1y---
thr a pretati m chs als cr; r.; c ssery t; stisfy t%; i,;:tu ti n ;.q;ii;.= :,t ,. The transmitters are located inside containment, with the taps in the vapor space region of the pressurizer, and thus possibly experiencing adverse , environmental conditions (LOCA. SLB inside containment, rod ejection). Therefore, the Trip Setpoint. reflects the inclusion of both steady state and adverse environmentet instrument uncertainties. l TheU,rj p~isetpoint"$0E51849?hJLfgyanja &bluleed
~
~$2TeGb 3.V$3 l This FunctibnmustbeOPERABLEin ES 1. 2. and 3 ;
(aboveP1f}tomitigatetheconseg aences of an HELB I o - inside containment. This signal may be manually i blocked by the operator below the 11 setpoint, j Automatic SI actuation below this ressure setpoint l .4 e , is then~ performed by the Contain t - g,% , Pressure-High 1 signal. 4M rn b ' "je f k - ,
$ 22-{a3 M 'I' j
~
1 (continued)[6
~ -
sr en MARK UP OF NUREG 1431 BASES B 3.3 87 5/15/97 i i
ESCAS Instrumentation B 3.3.2 L~ a BASES g)7.*H 3 APPLICABLE +1+ Steam Line Pressure - Low SAFETY ANALYSES, ,dTead)fu f-fianler (contin LCO, and Steam Line Pressure-Low mus be OPERABLE i l#M'l APPLICABILITY MODES 1, 2, and 3 (above P 1 when a
^- secondary side break or stuck ooen valve N
de A Sf8Na- j could result in the rapid depressurization of f/um / ,,, /re.rure-bw } the steam lines. This signal may be manually 6 n c h n Te > blocked by the operMr below the P 11
/ setpoint. Below P-; , feed line break is not a concern. Inside containment SLB will be f)?.743 terminated by automatic SI actuation via Containment Pressure-High 1, and outside-containment SLB will be terminated by the Steam Line Pressure-Negative Rate-High signal for steam line isolation. This Function is not required to be OPERABLE in s H0DE 4, 5, or 6 because there is insufficient energy in the secondary side of the unit to cxx en saidcat. M (2) Stc = Lin: Prc nurc !!ich Differcatici Irc::;;rc ".,-t.;;;; St=c Lin :
Stc = Linc Prcssurc !!igh Differcatisi Irc,surc Sct x n St G; Linc; picvidc; - protectica ascinst the fc11 swing accidents.
- SLS; ra d lin; brc;k; and p
InadVcitcat opening Of On 50 relief cr en 50 s;fety valvc. Stc = Linc Prcs urc lligh Siffercatici Prcssurc Sctwn n Stc = Lincs provides ac input to any catrol functicas. Thus, threc 0^CPASLC chenacis on cach stc = liac arc
. . . . - - sufficient to xtisfy the rcquirc.xnts, with a two cut of thru logic = cach stc= line.
b I b F b 5 e3H) bb b
. inside the stc= tunncis, it is possibic for
.. (continued)
MARK-UP OF NUREG 1431 BASES B 3.3-89 5/15/97 i l
ESFAS Instrumentation B 3.3.2 A =gP BASES APPLICABLE (1) Steam Line Pressure Lnw (continued) SAFETY ANALYSES LCO, and 1 i provides closure of the MSIVs in the event of APPLICABILITY a feed line break to ensure a supply of steam for the turbine driven AFW pump. Steam Line
/ Pressure-Low was discussed previously under 3
4 SI Function 1.e and!theHrip,3etpointlis sn Ty*eck on the[same bf*fa,. Iow rkJn'f*'"y Steam Line Pressure-Low Functi y y-f3 OPERABLE in MODES 1, 2, and 3 (above P l' . $ 770 with any main steam valve open, when a secondary side break or stuck open valve could result in the rapid depressurization of the steam lines. This signal may be manually g.pn,.h],f,ck, blocied by the operator below the P 11 U Terpoint.+ "#; " 11, an inside containment
/,,/,v[--Il)Oc SLB will be terminated by automatic actuation p L.
via Containment Pressure-High 2. Stuck T un m en A 4 g J,* valve transients and outside containment SLBs g(gg/)gl6. On will be terminated by the Steam Line Pressure Negative Rate High signal for Steam Line IIdCk' ) Isolation below P 11 when SI has been I manually blocked. The Steam Line Isolation Function is required in MODES 2 and 3 unless i all MSIVs are closed. er.d dc ;;tiv;ted. This l Function is not required to be OPERABLE in MODES 4, 5, and 6 because there is insufficient energy in the secondary side of cA.y,3.e,74
,,. .p.g the unit t#h;;; er. ;;;ider.t. M k h k 3 8 N# 8Y ##' 7 " (2) Steam Line Pressure - Neaative Rate - Hiah f an-} quijone,r} .
l Steam Line Pressure-Negative Rate-High ; provides closure of the MSIVs for an SLB when j less than the P 11 setpoint, to maintain at least one unfaulted SG as a heat sink for the ; reactor, and to limit the mass and energy release to containment. When the operator manually blocks the Steam Line Pressure-Low main steam isolation signal when less than , the P 11 setpoint, the Steam Line Pressure - Negative Rate-High signal is automatically j:.p (continued) MARK UP 0F NUREG 1431 BASES B 3.3 103 5/15/97 E
i ADDITIONAL INFORMATION COVER SHEET
' ADDITIONAL'INFORMATION NO: Q 3.3 APPLICABILITY: CA, CP, DC, WC REQUEST: [ Note 2 for ITS SR 3.3.1.2 is revised to replace the bracketed 12 hour time constraint with 24 hours. No such time limit is contained in the current TS, which has an !
SR.4.0.4 exception. The performance of ITS SR 3.3.1.2 should be provided sufficient time to ! perform the initial comparison and gain adjustment, equal to the SR frequency.] ; Note'2 for ITS SR 3.3.1.3 is revised to replace the bracketed 15% RTP power level constraint
' with 50% RTP. The specified power level in ITS SR 3.3.1.3 should reflect the applicable safety !
analysis basis consistent with the [ Applicability and] Required Actions of ITS LCO 3.2.3 (AFD) i l and LCO 3.2.4 (QPTR).
~ Comment: Reject - Scope issue - because ITS Note 2 to SR 3.3.1.2 and Note 2 to SR 3.3.1.3 proposes 24 hours for delay in performing the SR which are generic .
changes to STS that are not included in an approved TSTF. The justification is inadequste. Adequate justification would be the format is not usable, or the requirements represent unsafe practices or may result in an operational hardship. Note (DC] ITS markup missing "3.3-96" annotation. Reject 50% RTP in Note 2 to SR 3.3.1.3. FLOG RESPONSE (original): The justification for the proposed changes, originsily provided in JFD 3.3-96cwill be modified to include more detailed information conceming the application of the CTS 4.0.4 exception to these surveillances. The following has been added to JFD 3.3-96:
"These proposed changes address the current practices conceming the application of the CTS i 4.0.4 exception to various STS surveillance requirements.
For'DCPP, the plant-specific revision to SR 3.3.1.2 necessary to incorporate the CTS is described in JFD 3.3-47. For Callaway, CPSES, and WCGS, the proposed change to STS SR :
- 3.3.1.2, Note 2, is provided to incorpor.ste the SR 4.0.4 exceptio1 provided in the equivalent ;
CTS surveillance, contained in CTS Table 4.3-1, Note (2). The wrrent practice conceming the application of the CTS 4.0.4 exception is to defer the first required calorimetric measurement and the associated normalization of altamate power indications until a scheduled testing j
. plateau is attained duririg the post-refueling outage power ascension. Typical post-refueling ,
outage power ascens!ons include a power increase to approximately 30% RTP; although, the current practice of some plants is to achieve a power plateau nearer to 50% RTP. Using typical ' power ascension rates of 3% RTP/ hour,5 to 12 hours are required to reach the first scheduled power plateau above 15% RTP. Based on plant experience, after a specified power plateau is attained, approximately 2 hours are required to stabilize the plant and perform the calorimetric j measurement.=Two additional hours are required to make any necessary adjustments to the Nuclear instrumentation System [ ] power indications. The proposed 24 hour time frame also ! provides a reasonabh allowance during which the effects of equipment failures may be addressed. For example, if the plant computers are not available, some additional time is
. required to manually collect the data and perform the required calculations. The proposed 24-
hour allowance after increasing THERMAL POWER above 15% RTP provides a reasonable time to attain a scheduled power plateau, perform the required calorimetric measurement, and make any required adjustments in a controlled, orderly manner. Plant procedures provide guidance concerning the use of all available power indications [(e.g., NIS and loop AT)) during a power ascension. Any significant inconsistencies would be resolved through the performance of a calorimetric measurement. Based on the current, conservative operating practices and the fact that secondary plant calorimetric measurements are typically more accurate when performed at higher power levels, it is riot necessary to require a formal calorimetric comparison to be performed prior to exceeding 15% RTP. The note has sufficient limitations to ensure the plant is operated in a safe manner. Note 2 to STS SR 3.3.1.3 addresses the conditions required to perform the incore-excore
<.omparisons of AFD. The STS includes an allowance of 24 hours (within brackets) after cxceeding a specified power level for the first performance of this surveillance. The proposed change to the time allowance of STS SR 3.3.1.3, Note 2, would adopt the 24 hour allowance, but specify the power level to be 50% RTP. The change is predicated on the current practices conceming the application of the CTS 4.0.4 exception to the CTS (e.g., for Callaway, CTS Table 4.3-1, Note (3)). Current practice is to perform this surveillance at a power level below 50% RTP. Based on plant operating experience,24 hours is a reasonable time frame to limit operation above 50% RTP while completing the procedural steps associated with the surveillance in an orderly manner.
The deferral of the power level above which SR 3.3.1.3 is required to be satisfied is also consistent with current operating practices concerning the application of CTS 4.0.4. Due to such effects as shadowing from the relatively deep control rod insertion and, to a lesser extent, the dependency of the axially-dependent radialleakage on the power level, the refationship between the incore and excore indications of axial flux difference (AFD) at lower power levels , is variable. Thus, it is prudent to defer the calibration of the excore AFD against the incore AFD until more stable conditions are attained (i.e., withdrawn control rods and a higher power level). The AFD is used as an input to the overtemperature reactor trip function and for assessing compliance with ITS LCO 3.2.3, " Axial Flux Difference." Due to the DNB benefits i gained by administratively restricting the power level to 50% RTP, no limits on AFD are imposed below 50% RTP by LCO 3.2.3; thus, the proposed change is consistent with the LCO 3.2.3 requirements below 50% RTP. Similarly, sufficient DNB margins are realized through operation below 50% RTP that the intended function of the overtemperature reactor trip function is maintained, even though the excore AFD indication may not exactly match the incore AFD indication. [Further, current practice at many plants is to administratively reduce the I Power Range Neutron Flux - High setpoint until this and other surveillances (e.g., peaking factors) have been completed and it has been demonstrated that the plant is ready for sustained operation at full power in accordance with the design and licensing basis.] In summary, the proposed changes are more restrictive than the CTS, but are consistent with the current operating practices conceming the application of the CTS 4.0.4 exception to the r equivalent CTS surveillances. The proposed changes allow power ascensions and associated testing to be conducted in a controlled and orderly manner, at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be acceptable for subsequent use." FLOG RESPONSE (supplement): Based on feedback received from NRC staff during a telecon on January 12,1999, the Bases for ITS SR 3.3.1.2 and SR 3.3.1.3 have been revised to capture salient points from the justifications provided in the original response. J
l l ! ATTACHED PAGES: Attachment 9, CTS 3/4.3 - ITS 3.3 Enclosure 58, pages B 3.3-61 and B 3.3-62 l l l l l
l RTS Instrumentation B 3.3.1 t BASES . SURVEILLANCE SR 3.3.1.1 (continued) ; REQUIREMENTS Agreement criteria are determined by the unit staff based on a l combination of the channel instrument uncertainties including l indication and readability. If a channel is outside the cr1teria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. i The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal ! operational use of the displays associated with the LC0 required channels. SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS MQtadlegggg charel cutpt every 24 hours. If the j calorimetric exceeds the NIS G ) - cher.ral at st by Q""4 l
> 2% RTP the NIS is not declared inoperable, but R@LgMjggi )
phanneljgsges must be adjusted'm ' power! If the NIS ch;r.ral cut pt m cannot be properly adjusted, the channel is declared inoperable. Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS pogggd:#stRm$ charel =tst shall be adjusted cons 4 tent with the calorimetric results if the absolute difference between the NIS g g -harel stpt and the calorimetric is
>l t RTP py,yy
&i- 'he second Note clarifies that this Surveillance is required only if reactor power is 215% RTP and that 42 H hourp ge 4s allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are M#W inaccurate.+ Refenence]ROJgg}938g!!30p(IggggGRadggstnatMe) $3S$
83.N/ gap;osedwstr J cuan_spipavajagaturza The Frequency of every 24 hours is adequate. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Together these factors demonstrate the change in the absolute difference between NIS anu heat balance calculated powers rarely exceeds 2% in any 24 hour period. (continued) MARK-UP OF NUREG 1431 BASES B 3.3 61 5/15/97
INSERT B ?s.3-61 Q 3.3-96 The 24-hour allowance after increasing THERMAL POWER above 15% RTP provides a reasonable time to attain a scheduled power plateau, establish the requisite conditions, perform the required calorimetric measurement, and make any required adjustments in a controlled, orderly manner and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use. i l I
\
l l I ! i I i j
RTS Instrumentation
, -- B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 (continued) j REQUIREMENTS .
j In addition, control room operators periodically monitor l redundant indications and alarms to detect deviations in channel ) outputs. l SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel outout , every 31 EFPD. If the absolute difference is 2 34 St, the NIS l channel is 11 OPERABLE cj) y,7.gg, beck -for d. but must be readjusted. The esle ' p,p. ore o.f #e com <<1ron it ifferencer due -is changein core. gower //d'ildN'" 'SYthe^$1 ant [cIn 5fbe' properly readjusted, the channel is declared inoperable. This Surveillance is performed to verify the f(aI) input to the Overtemperature aT Function. Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between I
~~
( QT.M4
- Surveillance is required only if reactor power is a 154 50s RTP and that 24 hours is allowed for performina the first Surveillance after reaching ist M RTP. .2~X/ JEW 8 F,3-42 8M4 The Frequency of every 31 EFPD is adequate. It is based on unit operating experience, consider.ing instrument reliability and operating history data for instrument drift. Also, the slow changes in neutron flux during the fuel cycle can be detected during this interval.
SR 3.3.1.4 SR 3.3.1.4 is the performance of a TAD 0T every 31 days on a STAGGERED TEST BASIS. This test shall verify OPERABILITY by actuation of the end devices. The RTB test shall include separate verification of the undervoltage and shunt trip mechanisms. Independent verification l of RTB undervoltage and shunt trip Function is not required for I the bypass breakers. No capability is provided for performing g such a test at power. The independent test for bypass breakers u pg (continued) MARK-UP 0F NUREG-1431 BASES B 3.3 62 5/15/97 i
1 INSERT B 3.3-62 Q 3.3-96 I
' Note 2 allows power ascensions and associated testing to be conducted in a controlled and
- orderly manner, at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been ,
verified to be OPERABLE for subsequent use. Due to such effects as shadowing from the : relatively deep control rod insertion and, to a lesser extent, the dependency of the axially- l dependent radial leakage on the power level, the relationship between the incore and excore ;
- indications of axial flux difference (AFD) at lower power levels is variable. Thus, it is acceptable i
' to defer the calibration of the excore AFD against the incore AFD until more stable conditions l are attained (i.e., withdrawn control rods and a higher power level). The AFD is used as an {
input to the Overtemperature AT reactor trip function and for assessing compliance with LCO , 3.2.3, " AXIAL FLUX DIFFERENCE." Due to the DNB benefits gained by administratively j restricting the power level to 50% RTP, no limits on AFD are !mposed below 50% RTP by LCO { 3.2.3; thus, the proposed change is consistent with the LCO 3.2.3 requirements below 50% ;
. RTP. Similar1y, sufficient DNB margins are realized through operation below 50% RTP that the ;
l intended function of the Overtemperature AT reactor trip function is maintained, even though -I the excore AFD indication may not exactly match the incore AFD indication. Based on plant
~
i operating experience,24 hours is a reasonable time frame to limit operation above 50% RTP ! while completing the procedural steps associated with the surveillance in an orderly manner. j I I i i i~ l 1 g. d t
, +- - -
,- w
ADDITIONAL INFORMATION COVER SHEET
' ADDITIONAL INFORMATION NO: O 3.3-97 APPLICABILITY: CA, CP, DC, WC :
REQUEST: The NOTE for SR 3.3.1.6 of ITS 3.3.1 has been revised to state "Not required to be performed until 72 hours after achieving equilibrium conditions with THERMAL POWER ) 2 75% RTP". This is consistent with the current TS Functional [ Units 2.a and 7] surveillance i requirement of Table 4.3-1, as modified by Note (6) and current operating experience of l 72 hours for performing the SR. { Comment: Reject - The JFD statement regarding equilibrium cond;oons: "This is ; consistent with the current TS Functional (Units 2.a and 7) surveillance requirement of ! Table 4.3-1, as modified by Note (6)". This JFD does not appear to be consistent with CTS as the JFD states. The CTS note 6 only refers to incore-Excore calibration above 75% of RTP. The portion about " equilibrium" is added to the CTS Note (6) in Enclosure 2 as redline. : I Provide data to support the statement that operating experience shows 72 hours is l required to perform the SR. FLOG RESPONSE (original): For Callaway, CPSES, and WCGS, the justification for the proposed changes, originally provided in JFD 3.3-97, will be modified to include more detailed information about the current practices conceming the application of the CTS 4.0.4 exception to this surveillance. For DCPP, plant-specific justification is provided in JFD 3.3-06. For Callaway, CPSES, and WCGS, the following has been added to JFD 3.3-97: I i
'These proposed changes address the current practices conceming the application of the CTS 4.0.4 exception to STS SR 3.3.1.6.
1 The proposed changes to the STS SR 3.3.1.6 Note are provided to incorporate the SR 4.0.4 l exception provided in the equivalent CTS surveillance (e.g., see Callaway CTS Table 4.3-1, ) Note (6)). For CPSES and WCGS, the current practice conceming the application of the CTS 4.0.4 exception is to defer the multi-point incore-excore calibration until a scheduled testing plateau above 75% RTP is attained during the post-outage power ascension. Based on plant experience, it is necessary to perform this calibration at power levels at or above 75% RTP. During a post-refueling power ascension, it is usually necessary to control the axial flux , difference at lower power levels through control rod insertion. For example, at CPSES, it was l determined that a multi-point calibration performed well below 75% RTP resulted in excessive l incore-excore axial flux difference deviations at full power. The deviations were attributed to rod shadowing effects on the base flux map and, to a lesser degree, the dependency of the axially-dependent radial leakage on the power level. 1 To collect the data necessary to satisfy this surveillance requirement, it is first necessary to obtain equilibrium conditions at the appropriate testing plateau. A full-core " base" flux map is obtained. The control rods are then inserted into the core to initiate a swing in the axial flux difference (AFD). Several partial flux maps are then collected at various values of AFD. Based on plant operating experience, failures in the flux mapping system can result in extended data collection times. The plant is then retumed to a stable condition. The test data is then analyzed l l
and reduced, and the appropriate calibration data for the excore data is calculated. The excore detectors are then adjusted, as necessary. Based on plant experience, after equilibrium conditions are achieved at the specified power plateau, approximately 24 hours are required to perform the base flux map, initiate the required AFD swings, and collect the required data. The data is typically analyzed and the appropriate excore calibrations are completed within the following 24 hours. Based on plant operating experience, an additional time allowance of 24 hours is provided during which the effects of equipment failures may be remedied and any required re-testing may be performed. For Callaway, a similar sequence is used during the post-outage power ascension; however, a ' single-point incore-excore calibration methodology is utilized rather than a multi-point methodology. Since an AFD swing and partial flux maps are not required, the time for performing the flux map and collecting the required data is less than the 24 hours discussed above (typically 4 to 6 hours duration). Similar to CPSES and WCGS, the subsequent 48 hours allow for the performance of excore calibrations and provide an allowance during which . the effects of equipment failures may be remedied and any required re-testing may be l performed. [Further, current practice at many plants is to administratively reduce the Power Range Neutron Flux - High setpoint until this and other surveillances (e.g., peaking factors) have been completed and it has been demonstrated that the plant is ready for sustained operation at full power in accordance with the design and licensing basis.] In summary, the proposed changes to STS SR 3.3.1.6 are more restrictive than the CTS, but are consistent with the current operating practices conceming the application of the CTS 4.0.4 exception to the equivalent CTS surveillance. The proposed changes allow power ascensions and associated testing to be conducted in a controlled and orderly manner, at conditions that provide acceptable results and without introducing the potential for extended operation at high ; power levels with instrumentation that has not been verified to be acceptable for subsequent l use." l FLOG RESPONSE (supplement): Based on feedback received from NRC staff during a telecon on January 20,1999, the Bases ! for ITS SR 3.3.1.6 have been revised to discuss equilibrium conditions and to capture salient I points from the justification provided in the original response. ATTACHED PAGES: Attachment 9, CTS 3/4.3 - ITS 3.3 Enclosure SB, page B 3.3-63 \ _
~
~
;.Z __ .a.._. !
l RTS Instrumentation p 'B 3.3.1
' .7 l 't
- - BASES ~
SURVEILLANCE .SR 3.3.1.4 7(continued) REQUIREMENTS ~ .is inc15ded'in SR 3.3.1.14. The bypass breaker test shall include a local gggg shunt trip MW. A Note has been added to indicate that this test must be performed on the bypass breaker-prior to placing it in service. , The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It.is based on industry operating experience,. , considering instrument reliability and operating history data. f SR 3.3.1.5 Xyn . : sc . ' - N SR 3.3.1.5:is the performance of an ACTUATION LOGIC TEST. The JE SSPS l iil ested t every 31 day's on a STAGGERED TEST BASIS, using the s semiautomatic tester. The train being tested is placed in'the.
' bypass condition, thus preventing inadvertent actuation. Through
^
the semiautomatic tester, all possible logic combinations, with
, -and without applicable permissives, are tested for each protection functi. ~' 3 ,.1 'r .
M
~ c M;c - ' The Frequency of every 31 days u, on-a STAGGERED-TEST BASIS is adequate. It is based on industry
- operating experience, considering instrument reliability and .!
- - ~ operating' history data. !
SR 3.3 1.6 SR 3.3.'1.~6 is a calibration of the excore channels to the incore channels. If the measurements do not agree. the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f(aI) input to the Overtemperatur T Function. Detegnfnattbag!@lhe9 pop 35pegggc - ns.= _ET*1 jdlbeimade3vhempegormingithisWl'1bratgefCAD-g7 i m g g conditjons2 4T, ten atsjlpriu1T
.... A Note modifies SR .3.1.6. The Note states t his- N '
Surveillance is required only if reactor power is Wig 5% RTP and that 24 72 hours afterfjacMeving;_eq@l1brJtselcondjMonsig_th 2 5,. THERMA 17. POWERJiUSE_ RTE is allowed for rformingthefirst surveillance. Ofter rscMr.;; 50% RT". T/i/erum ed/8mr an 89.347
.rdelle af in&ed echiewd when f(e con i.e .rufViereor 0/**=h cad /1'fmr /r/'M f/ux yf fy. (continued)
ZNMRT~ B 7,3'43 y MARK UP OF NUREG 1431 BASES B 3.3 63 5/15/97
INSERT B 3.3-63 Q 3.3-97 i The SR is deferred until a scheduled testing plateau above 75% RTP is attained during the l post-outage power ascension. During a typical post-refueling power ascension, it is usually necessary to control the axial flux difference at lower power levels through control rod l insertion. After equilibrium conditions are achieved at the specified power plateau, a flux map must be taken and the required data collected. The data is typically analyzed and the : appropriate excore calibrations completed within 48 hours after achieving equilibrium conditions. An additional time allowance of 24 hours is provided during which the effects of i equipment failures may be remedied and any required re-testing may be performed.
- The allowance of 72 hours after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascensions and associated testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without !
introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use. t l t 1 l I !. i j i { r i 4 i 4 j I i t l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CP-3.3-012 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.3.1 Bases are revised to clarify the relationship between the Reactor Coolant Flow - Low reactor trip function (Function 10 in ITS Table 3.3.1-1) and permissives P-7 and P-8. Traveler TSTF-169 correctly combined NUREG-1431 Functions 10.a and 10.b into a single Function 10. A single function is appropriate to ensure thermal power is reduced below the permissive P-7 setpoint in a time frame consistent with WCAP-10271 and its supplements. However, the Bases changes attached to TSTF-169 did not adequately revise all discussions of this trip function. In addition, as noted in a telecon with NRC staff on January 12,1999, it is confirmed that the monthly Actuation Logic Test of ITS Table 3.3.1-1 Function 21 (ITS SR 3.3.1.5) demonstrates how the low flow trip function on low flow in two or more loops and on low flow in one loop is enabled or blocked by the presence or absence of permissives P-7 and P-8, respectively. ATTACHED PAGES: Attachment 9, CTS 3/4.3 - ITS 3.3 Enclosure SB, pages B 3.3-23, B 3.3-24, B 3.3-37, B 3.3-38, B 3.3-52, and B 3.3-53 l l a
RTS Instrumentation
,g B 3.3.1 Q11: ,
BASES APPLICABLE 9. Pressurizer Water Level - Hiah (continued) SAFETY ANALYSES._ LCO,'and fourth channel is not required to address control / ' APPLICABILITY protection interaction concerns. The level channels do not actuate the safety valves, and the high pressure reactor trip is set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip. l In H0DE 1, when there is a potential for overf1111ng the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P 7 interlock 20n decreasing power, this trip Function is automatically blocked below s P 7. Below the P 7 setpoint transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and
-take corrective actions.
- 10. Reactor Coolant Flow - Low
- e. :::t;- C=t . n = L = 'Si = k L _.)
b ~ N' The Reactor Coolant Flow-Lowd5GUE63D trip QCy,7-o/p. O r**
- O M r * /* "8 N " '
Function ensures that protection is provided against. In b 8 d r N r* A C- violating the UNDK limit due td' low flow in one or l enMficalf twM . 7' '# more RCS loops, while avoiding eactor trips due to normal variations in loop flow.WAbove the P 8 setpointCWFuctu1r _; ; ==Auuga loss of flow in any RCS loop will actuate a reactor trip. '"' Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control l system input. lSW N The LCO requires thre Reactor Coolant Flow-Low cg g,y 0/2 channels per loop to be OPERABLE in H0DE 1 abov A7 (twointf9Ethteenw 3 = pItc E.tosic E M !B W i Setpottt.11 G I90E oG_ cop?pjjatqReashL'WITR8 l 09tjF $951660igpag vin MODE 1 above the P 8 8 S N*^~ ! i setpoint, a loss of flow in one RCS loop could .
- w. result in DNB conditions in the coreg In MODE 1 08-334/2
- h, kure rP }he hojkew fewee lenI.
(continued) MARK UP OF NUREG 1431 BASES B 3.3 23 5/15/97 L
a~.-.e..e...--..,-.e, RTS Instrumentation _. B 3.3.1 BASES v APPLICABLE er Reactor Coolant Flow - Low 'Sincic Lad ANALYSES, (continued) {p, APPLICABILITY. below the P 8 setpointy a loss of flow in two or more loops is required to actuate a reactor trip (Tunction 10.b) because of the lower power level and 8e/,w f4e 47.reporn-h ,// the greater margin to the design limit DNBR.
\ A reachiny#
878 delwna ca ltylon low.. Abw
//eckal "=ctor Cccl=t rica Lcw 'Tuc L x d '
?Inee. f$ece ir ifnru(frered hg g/,,fg 4 Reactor Coolant Flow-Low (Two Loops) trip
.Im4 g Q unction ensures that protection is provided aga st y lating the DNBR limit due to low flow in or mor RCS loops while avoiding reactor trips e to norm variations in loop flow.
Above th. P 7 setpoint, ;nd bclew th: 0 ;ctp; int, a loss of ow in two or more loops 11 initiate a reactor trip Each loop has three low detectors to monitor flow. The flow signals e not used for any control system t. The LC0 requires th Rea or Coolant Flow-Low channels per loop to LE 32EingwfEtheire inlagdoopitgip$i,,pg-MEsnyg16Eis s190TigfHepNJ- _ +1W0mfX95Ei!!O gpag In H0DE 1 above e P 7 setpoi and below the P 8 setpoint, the actor Coolant F ow (Two Loops) trip must be PERABLE. Below the 7 setpoint. all reactor tri s on low flow are autom ically blocked N"E"..IMII! k[I! !! NN5s bI E M5bnst[fhe}hi5a}hph5ctY [o!i>e conc _ned!about[DNB. Above the P-7 setpoi , the re tor trip on low flow in two or more RCS oops is tomatically enabled. Above the P-8 setpoin a oss of flow in any one loop will actuate a re tor trip because of the higher power level and the reduced margin to the design limit DNBR. v (continued) MARK-UP 0F NUREG 1431 BASES B 3.3 24 5/15/97
- . .. . - . ~ =. - . - . - . - - _ .
RTS Instrumentation ; q B 3.3.1 ' %M < w - BASES' APPLICABLE- b. Low Power Reactor Trios Block. P (continued) ! SAFETY ANALYSES, LCO, and (1) on increasing power, the P 7 interlock ; APPLICABILITY- automatically enables reactor trips on the ! (continued) following Functions: Pressurizer Pressure-Low:
.. Pressurizer Water Level-High:
i C8:LT-#/2 Reactor Coolant Flow-Low ( :
.....,.,,.,,,,-m,, r ouflew i
. Undervoltage RCPs: and rn ym, ,,
New Mcs s . Underfrequency RCPs.
/ufs These reactor trips are only required when ,
operating above the P-7 setpoint ' (;.pp existely 10t power). The reactor trips provide protection against~ violating the DNBR limit. Below the P 7 setpoint . the RCS is capable of providing sufficient natural , circulation without-any RCP running.- l (2) on decreasing power, the P 7 interlock ' automatically blocks reactor trips on the following Functions: Pressurizer Pressure-Low:
. Pressurizer Water Level-High:
. Reactor Coolant Flow-Low ( .): ~ #* N
. P,CP Sraict Positicr. 'T e LeopO;
{
. Undervoltage RCPs: and
. Underfrequency RCPs.
(continued) j MARK UP OF NUREG 1431 BASES B 3.3 37 5/15/97
- m. _ ._ - _ . _ 1
. h 4 . .- .
RTS Instrumentation l 4 B 3.3.1 1 ' ' ?W1. :p .
' BASES i APPLICABLE' b. Low Power Reactor Trios Block. -P-7 7(continued) ,
SAFETY ANALYSES, . I
- LCO, and Trip R,,eir.t; .c4 A110wable' Values are not 'l APPLICABILITY. , applicable to the P 7 interlock because it is a l
. logic Function and thus has no parameter with which -
to associate an'LSSS. l The P 7 interlock is.a logic Function with train and not channel . identity. Therefore, the LCO requires one channel per train of Low Power Reactor Trips i Block, P 7 interlock to be OPERABLE in MODE 1. I
~
The low power trips are blocked tielow the P 7 . setpoint and unblocked above the P-7 setpoint. In - ! MODE 2, 3,l4, 5, or 6. this Function does not have to be OPERABLE because the~ interlock performs its ( -Function when power level drops below 10% power, which is in MODE 1. O c. Power Ranae Neutron Flux. P 8
'The Power Range' Neutron Flux, P 8 interlock is lg actuated at .p,,rexi teli 48% power as determined by I two out of-four NIS power range detectors- The P 8 interlock automatically enables the Reactor Coolant Flow-Low d5iiE305BB.r :" =. "rnter ";;iti=- Cf .r.3-v/2 !
Cir.;1c Ler+; reactor trips.on low flow in one or
- more RCS loops on increasing power. The LCO
^
l requirement for this' trip Function ensures that protection is provided against a loss of flow in any RCS loop that could result in DNB conditions in the 1 core when greater than apprexi n tcly 48% power. On decreasing power, the reactor trip on low flow in any loop is automatically blocked. The LCO requires four channels of Power Range ! Neutron Flux, P 8 interlock'to be OPERABLE in MODE 1 2 (twoloutspfifouttfdM_N4appoigds M38tXU In MODE 1. a loss of flow in one RCS loop could result in DNB conditions, so the Power Range Neutron Flux, P 8 interlock must be OPERABLE. In MODE 2, 3, (continued) MARK-UP 0F NUREG 1431 BASES B 3.3 38 5/15/97 t _ _
V w , - . . . -..m... _ _ . - ~ . . _ . -- i RTS Instrumentation j
- g. B 3.3.1 !
1 J
;r gt Ai i
l BASES ACTIONS L.1. L.2. and L.3 (continued) _..____;_; ,__;,.._,.. ____,.L dudysarusu .- wv.ubsig. w2_
.ned .2,.,.
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.._L. .s J ..L_ - - . . _ - _ _ . 211 ____1..J. - L____ 221..s2__ ___2a__t us as vi u ww%s seu ww s dvui ww a vsaau yswwawww u vvi m a u a s u w s vi s uww a wws a b .
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\ ubs%s a b # vi ud , b3 rws w I v3 wa wvs w awwwb5y aby du evu u %s a sv b ww IIist bu d 5 6 sy ,
__J _ 6 L..._ -.J-_..m. - A2._ _r u, u u 1.h ,~m r s_w y_,.._-
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s ~_ vv e r-1..bsm. 1-2. s -d_ _ v.
.2&L2 1 L..._ __J ____ ___ go L_..__ ___ L-__; __ ____ s2__
yv 3 b u s i s a A u rvus un r%s VI rww yws 45. a rvus d us w wudw%s VIB vys a u b a u ry I _.____2____ J .- ___f__2__. AL_ n__.J__J adJ-__ -J LL. L._ .1 J-_ l wisywu s ws sww ses ybs sv3vussary wars s%w3u u s wu nw k s vz ud uuvu bars rsa rvvu s wuys i sL.A . ._2 m ___;2A2___ . 211 .L..___ _1_ .1 . . ba m b ma i b wm m a b s .. .d v. s vs ,m ,vs e.-.g. I M.1 and M.2 - 1 Condition H applies to the following reactor trip Functions:
. Pressurizer Pressure-Low- i
. Pressurizer Water Level-High; C[-3.Wd / 2.
7 _e.t - .L. oo _ .-
~
A, .:- d:+ _,2 7 wymy ued
. Reactor Coolant F1ow-Low 6: !1 I
I n,~rn. n__.t__
,, ,____s.
. . , - . n _ u. 2 m_ _a
,ed r,.,,._v vvyals
. Undervoltage RCPs: and l
. Underfrequency RCPs.
With one channel inoperable, the inoperable channel must be !
, . placed in the tripped condition within 6 hours.gRMJV(lie 3 ZNrF. EXT ~ f.T.2-D- Y (continued)
MARK-UP 0F NUREG 1431 BASES B 3.3-52 5/15/97
(
.1 l
l I INSERT B 3.3-52 CP-3.3-012 i 1 l For the Pressurizer Pressure - Low, Pressurizer Water Level- High, Undervoltage RCPs, and Underfrequency RCPs trip Functions, placing the channel in the tripped l condition when above the P-7 setpoint results in a partial trip condition requiring only one ! additional channel to initiate a reactor trip. For the Reactor Coolant Flow - Low trip i
. Function, placing the i i
i 4 I l l i l 1 l 1 l o ! i L i l
_ _ _ _ . _ _ _ ._ . .-_~__ . _ _ _ -_. . _ . _ - - . ._ RTS Instrumentation B 3.3.1 BASES I 7C ( When *Leve die ACTIONS M.1 and M.2 (continued) g y fgg-( fn +4e .rerne lco,o g channel in the tripped condition #esults in a p 1al trip condition requiring on1 one additional channel to initiat_e a 7[.-
# " b reactor trip o
.c ',o"n ~u,-.
/C r los,M Ia/M
8g * /r. -.m ;;tpcint.4 These Functions do not have to ERABLE a reuder Njo w/en below the P 7 selpoint because there are no loss of flow trips
/ ,/, 9 .,c4 , je. f .below the P 7 setpoint.d The 6 hours allowed to place the channel 444j,, ,
in the tripped condition is justified in Reference 7 5. An ff # 'jf. onal 6 hours is allowed to reduce THERMAL POWER to below h* 7 if the inoperable channel cannot be restored to OPERABLE status or blaced in tri within the specified tion Time. W ere Tc j
. , - g. .._....._.7
' beaffrvlue}Gn bgenern/a hNg , C88 /Y/'"' Allowance of this time interval takes into consideration the ! le /ew +Ae 47 redundant capability provided by the remaining redundant OPERABLE l telj9/"rd i channel, and the low probability of occurrence of an event during ! this period that may require the protection afforded by the Functions associated with Condition H. i The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 4 hours while performing routine surveillance testing of the other channels. The 4 hour time limit is justified in Reference 7 5. N.1 and N.2 NotiUsedi. Cendition " applic; t; the 3;;ctor C001;nt flow-L;w 'Singic L;;p) reacter trip ranction. With ca; ;M anci ir.;perabic. the inspcretic chonnci au;t h pleccd in trip within S h;;r;. If th; chenaci connet b; rc;tered to OP:Pa LE ;tatu; or th; ;Manci ; pieced in trip within th: 5 hour;. then IlEP"."L ""'ER ;u;t bc i reduced below the P-3 ;;tpcint within th; n;xt 4 hour;. Thi;
.3, piecc; the unit in "00C J. crc the LOO i; no lenscr opplicabic.
i' (continued) i MARK UP OF NUREG 1431 BASES B 3.3 53 5/15/97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.11-4 APPLICABILITY: CA, CP, DC, WC REQUEST: Change 4-09 LS-36, Difference 3.4-47, Change 3-04 and Difference 3.4-31 Comment: WOG-87 has not yet become a TSTF. FLOG RESPONSE (original): As discussed during a telecon with NRC Staff on July 30,1998, the above references to DOC 3-04 and JFD 3.4-31 apply to NRC-approved traveler TSTF-87 and were not intended to be questioned here. Additional changes have recently been added per Revision 2 of WOG-87 and are included in the attached pages below. The addition of the Note to the block valve Action Statement is considered to be administrative in nature as it reflects current plant practice. WOG-87, Revision 2, has been approved by the TSTF group and is expected to be submitted to the NRC expeditiously. Given the nature of the Notes added to the PORV block valve Required Actions and Surveillance Requirement, the FLOG continues to pursue the changes proposed by this traveler. FLOG RESPONSE (supplement): WOG-87, Rev. 2 was approved by the TSTF and designated as TSTF-309. As discussed with the NRC staff on January 19,1999, the NRC agreed to approve TSTF-309 for the FLOG with a clarifying Bases change conceming placing the PORV(s)in manual control. ATTACHED PAGES: - Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure SB, pages B 3.4-58 and B 3.4-60
Pressurizer PORVs
' B 3.4.11 6 /;j Ihg BASES ACTIONS B.I. B.2. and B.3 (continued)
If one or-4ne PORVt is inoperable and not capable of being manually cycledg it must be either restored or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of I hour are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restored to OPERABLE status, it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status. If the PORY cannot be restored within this additional time, O pl .; ; ;; b; trCf. t; ; F-1 ir did tra LOO t . /. ; y, n .;.;,1.-d by Condition D,648U/":3
, dmuslehen C.1 and C.2 f M3w W10 KL <yh}- 6~M 5 d feaA r re[m If one block valve is inoperable, then it is necessary to either restore the' block valve to OPERABLE status within the Completion Time of 1 hour or place the associated PORY in manual control.
# The prime importance for the capability to close the block valve is to isolate a stuck open PORV._ Therefore, if the block valve cannot be restored to OPERABLE ' status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of I hour is reasonable, based
. on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. Because at least one PORY remains OPERABLE, the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status. The time allowed to restore the block valve is based upon the Completion Time-for restoring an inoperable PORV in Condition B, since the PORVs are not - capable of mitigating an n;.pc;;; re event when plee d ir. c ;1 e tcel. o-If the block valve is restored within the Completion Time of 72 hours, t.'a pn;r d11 b; r;;t;,rd ;d the PORY
~
r;;tc,c d t; 0 O LE ;t;t ;. ~ - M If it cannot be restored within this additional time,
.g ( tr.; pl r.t ; ;t b; b. 4t te ; "00C ir did it.; LOO t.;,;; r.;,t n .;,,;ird t# tondition D, M d2 3.t.//-3_
~
,.,.1
*Y ll ZWSERT$AfST (continued)
MARK UP OF NUREG 1431 BASES B 3.4 58 5/15/97 L
~':
INSERT B 3.4-58 Q 3.4.11-4
.Q jy The Required Actions are modified by a Note stating that the Required Actions do not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition. While it may be desirable to also place the PORV(s) in manual control, this may not be possible for all causes of Condition B or E entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse (s) or control switch malfunction (s)).
l l s d b
Pressurizer PORVs B 3.4.11
> f. .c qq,v,f ACTIONS E.1. E.2. E.3. and E.' (continued) to reach the required plant conditions from full power conditions in orderly manner and without challenging plant systems. In f+jc-f MODES $8Mb/ l
- a tteir.ir.; M PORY OPERABILITY a may-be required. See LCO 3.4.12.
r- T l F.1, F l.
.-d ".2 N 8d *"'I
.-U (rdline)
,If more than one block valve is inoperable, it s necessary to either restore the block valves within the Coup etion Time of 1 hour.. or place the' associated PORVs in manual control and ,
restore.at least one block valve within 2 hour .74 r;;O.; th;$f.44en-/
. 217.; t'ai sk; i.th;r. 72 har;.--- The Completion Times are '
reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation. ~ 1hlft h A ?.4-1(0 $ NNA us.muummma M ; If the Required Actions of Condition F are not met, then-the
;.1;;.; ;t M L;zf.; O ; % 17. .i.i2. ;M L;0 C;.; ;~; ,,,.ly.
i T; ahkx thu ;t;t ;. the plant must be brought to at least l MODE 3 within 6 hours and to i MBBE-4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. InMODE% o . . -$ FA6,m/
~
M 2 3teir.ir.; -; : PORV ILITY sey-be 3 required. See LCO 3.4.12.
-,1 Ie nna 9 (continued)
MAPK.UP OF NUREG 1431 BASES B 3.4 60 ~ 5/15/97 i
INSERT B 3.4-60 Q 3.4.11-4
, -- s l
"%$l The Required Actions are modified by a Note stating that the Required Actions do not apply if j the sole reason for the block valve being declared inoperable is as a result of power being l removed to comply with other Required Actions, in this event, the Required Actions for - !
inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition. While it may be desirable to also place the PORV(s) in j manual control, this may not be poss.ible for all causes of Condition B or E entry with PORV(s) ; inoperable and not capable of being manually cycled (e.g., as a result of failed control power ' fuse (s) or control switch malfunction (s)). , i e t f s r p i I l l s >
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.12-2 APPLICABILITY: CA, DC, WC REQUEST: Differences 3.4-23 and 3.4-45 Comment: WOG-51 Rev.1 has not yet become a TSTF. FLOG RESPONSE (original): WOG-51, Rev. 2 is designated as TSTF-285. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-285 was modified from WOG-51, Rev. 2, and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes proposed by this traveler. FLOG RESPONSE (supplement): As discussed with the NRC staff on January 19,1999, the NRC agreed to approve TSTF-285 for the FLOG with aoditional Bases changes to provide an explanation of the time and tasks associated with the charging pump swap evolution. ATTACHED PAGES: 0, CTS 3/4.4 - ITS 3.4 - 8, page B 3.4-71
r anw eje m
'I B 3.4.12 BASES APPLICABE RCS Vent Performance SAFETY ANALYSES (continued) With the RCS'depressurized, analyses show a vent size of EM square inches-is capable of sitigating the M eHeved4-10P M overpressure transient. The capacity of a vent this size is greater than the flow of the limiting transient for the b10P -
configuration, a; ll": 77 ;.2 - - OPERABE, maintaining RCS pressure less than the maxima pressure on the P/T limit curve. The RCS vent size will be re evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel . material surveillance. The RCS vent is passive and is not st& ject to active failure. The L T Oy;ts satisfies Criterion 2 of t.t = ";M;y Statemen6-LCO This LC0 requires that the LW Sy;ts is OPERABE. The L"" Sy;ts M is OPERABE when the sinteunt coolant. input and m pressure relief capabilities are OPERABE. Violation of this LCO could lead to the loss of low temperature
~
overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient. To limit the coolant input capability, the LCO requires
;,.; ll"I m m capable of injecting into the RCS and all accumulato ischarge isolation valves 3 closed and gyg"~j
" - Tamobilize d ,
accumulator pressure is greater than .e equal to the maxi RCS pressure for the existing RCS cold leg temperature allowed in the PTLR
,, OMY'[Q cA-3.4-x 2.
Q 3.'+./2 .2 l
}f. W, .
/ houe f12.+, l.2-S (continued) f MARK UP OF NUREG 1431 BASES B 3.4 71 1 5/15/97 W
8 2.+-7) $ 3.+./2-2 L- - - - -
' ~ INSERT B 3.4-71 Q 3.4.12-2
~,.; ,
%l:N
- One hour provides sufficient time to safely complete the actual transfer and to complete the administrative controls and surveillance requirements associated with the swap. The intent is to minimize the actual time that more than one centrifugal charging pump is physically capable of injection.
6 9 4
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6
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m i . .
. i . . . . . .
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.6.3-54 APPLICABILITY: WC, CA REQUEST: ITS B3.6.3 Bases - C.1, C.2, and C.3 Based on the changes made to ITS 3.6.3 with regards to purge valve leakage testing, a number of sentences in ITS B3.6.3 Bases - C.1, C.2, and C.3 which refer to SR 3.6.3.7 also need to refer to SR 3.6.3.6. These sentences are first paragraph, fourth sentence; third paragraph, first sentence and third paragraph, third sentence. Comment: Revise the ITS markup accordingly or p.rovide a discussion and justification to show that the reference to ITS SR 3.6.3.6 should not be added. FLOG RESPONSE: (original) ! The ITS Bases have been revised to include reference to SR 3.6.3.6 as well as SR 3.6.3.7 in the three locations mentioned in the comment. The Callaway markup already had added SR 3.6.3.6 in two of the three locations. l FLOG RESPONSE: (supplemaat) ; J In Reference 1, a supplemental response was provided to Comment Number ; Q 3.6.3-21. Based on the additional changes proposed in the supplemental response, i the ITS B 3.6.3, ACTIONS C.1, C.2 and C.3 should have been revised to delete "SR 3.6.3.6 or" that was added in response to this Comment ivumber. This Comment Number is using revised to reflect the change proposed in the supplemental response to Comment Number O 3.6.3-21. ATTACHED PAGES: Att. No.12 CTS 3/4.6 -ITS 3.6 Encl. SB B 3.6-24 i i 1 1 I i k-
Containment Isolation Valves (At;r.;;phcric. Subtresphcric. Icc Ccadenscr ;nd Duci) B 3.6.3
/@?%
BASES 7%
/ x ACTIONS - 6 E-1-E-e-end-E-3 ged) pg/n-84 V
Action does not or valve anipulation. verificrequire any diontestTng/ f Rather. it involves , thicagh a systc; walkdc.;ns that i those isolation devices o ide containment capable of being / mispositioned are in the c rect position. For the isolatic devices inside containment, he time period specified as " pior , to entering H0DE 4 from 5 if not performed within t previous 92 days" is bas on engineering judgment an s 'd' ! considered reasonable in view of the inaccessi of the 1 [ I isolation devices and ot r admini ensure that isolation dev~ ntrols that will gnment is an unlikely b possibility. mig (sygf m pgkJ. ,de atc / focal rwok ink &eM) l
' ~
For the.gontainment alve'Mth-resirh { seal that is isolated in accordance with Required Action ; ,
)
s E.1. m SR 3.6.3.7 must be performed at least once l every 3 days. This assures that degradation of the resilient y seal is detected and confirms that the leakage rate of the /f 5, containment purge valve does not increase during the__ time the - -Q
% penetration is isolated. The normal Frequency forgR 3.6.3.7 ^7.;;.g 184 days, is based on an NRC initiative. .c ' . . l.~1 & s Z; g g,y.sy l
eJ8 B 20 (Ref. 3 g). Since more reliance is placed on a single
--' ~
valve while in this Condition. it is prudent to perform the SR more often. Therefore, a Frequency of once per E days was /
@ 3,ylls- chosen and ha? been 2.:r.n to be acceptable based on operating i/Q 3.G.0 -2.7 W
g 'D.'L- experience. y -_
} -
l
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tne.,_segngismfffledN75)!!tyg.c51rMporttichTpKo!fito IschhKdifaltggxoEswring (continued) I i MARX-UP OF NUREG 1431 BASES B 3.6-24 5/15/97 : 4
J l ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: CA-3.6-ED 1 (New) APPLICABILITY: CA i REQUEST: Editorial change to be consistent with previous change sent under CA 3.6-001. The reference 4 change is also being reflected in the text of Bases SR 3.6.1.2. l l ATTACHED PAGES: Att. No.12 CTS 3/4.6 -ITS 3.6 Encl. 58 8 3.6-5 l l i
1 Containment (At;;.e:p,;ri;) ! B 3.6.1 BASES 1 d SURVEILLANCE SR 3.6.1.1 (continued) REQUIREMENTS s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as , required by A f , Radix 0, es ;. edified tg ;- m rd c ---;.ien; the C" si=#"NN#dtCTfstT ! 6-lih= Tf;q ui['deneicas; e decs ..vi o wl, M periuuicT; e;. ';; 3.v.e (wni testing requirements verify that the containment leakage rate
] -
does not exceed the leakage rate assumed in the safety analysis. 59- 3.6.1.2 Tor unsre t;d pst t;n;iened t;nden;. This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon ("""ei'l?"^a Dea 0"e= lestia; a~J Caaq"=acy are caashtant with " 3M)$q) kk ,e rec- = detiena vi RevuiaLucy Guiue 1.33 (Rei. N. A/- - -y n a 3 hr,-o REFERENCES 1. 10 CFR 50. Appendix JHip[jorrJ.
- 2. FSAR. Chapte 15.
- 3. FSAR ction 6.2. '~
,. .n,-.,o.ouviy m
uuiue 2.ss.
, ~ . . . . . .
+v y FSAKC/@ lG.
5- "!?M^* r 5"*" ' '* m-- --'LM i CA 3 C-001 J l MARK UP OF NUREG 1431 BASES B 3.6 5 5/15/97
cr
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I ADDITIONAL INFORMATION COVER SHEET ! ADDITIONAL INFORMATION NO:' Q 3.7.16-3 APPLICABILITY: WC, CA , I
. REQUEST: ' CTS 3.9.12. Action b (for Callaway and WCGS)
STS 3.7.16 -
]
JFD 3.7-dd STS 3.7.16, Fuel Storage Pool Boron Concentration, was not adopted. ) 1 Comment:. 'There is no detailed explanation for why this STS LCO was not adopted. > Also, see Comment 3.7.17.1-2 of ITS 3.7.17. j FLOG RESPONSE: ITS 3.7.16 is a bracketed specification in' the STS and per the STS conversion methodology would not be required to be adopted unless it already existed in the CTS (as was the case for DCPP). Callaway, CPSES and WCGS do not have this specification in their CTS. For CPSES, the NRC SER for the high density rack License Amendment (LA 46/32) concluded that nonnal plant procedures were sufficient to assure boron concentration in the spent fuel pool. Callaway and WCGS are further evaluating STS 3.7.16, Fuel Storage Pool Boron Concentration and CTS 3.9.12, Action b, in conjunction with the CTS license amendment request for reracking of the Spent Fuel Pool. An additional response to Comment Number Q 3.7.16-3 and Q 3.7.17.1-2 will be provided by November 6,1998. FLOG SUPPLEMENTAL RESPONSE: Callaway and WCGS intend to adopt STS 3.7.16, " Fuel Storage Pool Boron Concentration." Additional analysis was performed to determine the value for the boron concentration based on CTS license amendment requests for reracking of the Spent Fuel Pool. Callaway was issued Amendment No.129 on January 19,1999, and WCGS expects to receive an amendment in February. 1 The mark-ups to STS 3.7.16 are based on Amendment No.129 for Callaway and the expected WCGS amendment. Remaining changes associated with Amendment 129 , and the expected WCGS amendment will be incorporated into the conversion application in the next follow-up letter. ATTACHED PAGES: Att. No.15 CTS 3/4.9 -ITS 3.9 ,' i
. Encl. 2 - CTS 3/4.9 new LCO 4
Encl. 3A' 10 L Encl. 3B 8 i Att. No.13 CTS 3/4.7 -ITS 3.7 Encl. 5A NUREG-1431 Specifications That Are Not Applicable Table, new LCO 3.7.16
' Encl. 5B new LCO 3.7.16 Bases -
I r}S.krk -- % o L C O ruei storage Pool Baron Concentration 3.7.16 CTS 3/4,1 (8/7fd Fuel Storage Pool Boron Concentration J4-09-M Q MW Q 3.'7. l f,- 3 LCO 3.7.16 The fu.t]. storage pool baron concentration shall be hI ] ppm. to APPLICABILITY: W en uel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the-last movement of fuel assemblies in the fuel storage pool.
' ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME-C A. Fuel storage pool ----------- NOTE---- ==-
boron concentration LCO 3.0.3 is not applicable. not within limit. ----------------- =--- A.1 Suspend movement of Immediately ., fuei assemblies in the fuel storage L pool. 8.ND A.2.1 Initiate action to Immediately restore fuel storage pool boron concentration to within limit. DE A.2.2 Verify by Immediately
' ag;inistrativemeans
[.._ ia. 2] fuel-gw storage pool ()* gb im { verification has been performed since the last movement of fuel assemblies in the fuel storage pool.
l Fuel Storage Pool Baron Concentration 3.7.16 "N {[ SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 ' Verify the fuel . storage pool boron 7 days. concentration is within limit. l l 4
\
g [
. 2-i t
l l . . l
, _ . . --- . = - - - . . . . - - _ .
E 1 l CHANGE
. NUMSER 85dC DESCRIPTION
,m ,
' operations. This change is consistent with NUREG 1431.
I
.,<.M- ,. ,-
Rev. 1. 14-04 LS 13 Deletes the action statement requirement to verify spent ! fuel pool boron concentration every 8 hours while action
~
is being taken to relocate noncomplying spent fuel
- assemblies from Region 2 to Region 1. This change is consistent with NUREG-1431. Rev. 1.
.Tdsh [ [h_-IDAL } lb '3.7. t7. t-+ l j
^
~~ "14 05 LG qu;,remen yow so e ou an ) 1 l' 1 ass lie in Re on 2 uld b relo tedtfa 1 cens . co rolled doc t. s cha ,e is nsis nt f
.f ith UREG- 431. v. 1. nd mo es re ireme s tha do
. t i...
~
-.t . _ ,.". '_ .J .. .
+ +.. ,. ,- fr nci '
the 14-06 A Not applicable to Callaway. See Conversion Comparison , Table (Enclosure 3B). l
'.;14 ,07 g f- - _ _ _ _ _ .Us&c[yA[/o/} Q 3.7./7. /-d j
14-08 / tG. u d/ -n/fst.r 7A /Oc - Io J7 /7M f _ - wlErr 3A ~ /003 ] 7.1 3l) 14-09 4ijSfd 7 , , g_, 7 d 14 10 A i. staf .,en*'thatg.0.4flis no appli/a'ble [ dele /ed. is change s consiste t wit NUREG-1431./Rev.1/. Th's hanp6 doe not esul in a hange o technical / ec,dir . ts. 15 01 R Not applicable to Callaway. See Conversion Comparison , Table (Enclosure 3B). l i 10 5/15/97 DESCRIPTION OF CHANGES TO CURRENT T5
p: i. Q 3.7.16-3 INSERT 3A-10e . 14-09 M A new LCO is added to CTS 3/4.9 to impose limitations on fuel
. storage pool boron concentration. This is considered to be a more restrictive change since it requires a minimum boron 1 concentration to be verified prior to moving fuel in the fuel storage pool.
b a l t [ i i s , i I i l ) L
-)
i I i l 1 1 1 I I l
CONVERSION COMPARISON TABLE - CURRENT TS 3/4.9 Page 8 of 8 TECil SPEC CilANGE APPLICABILITY NUHBER DESCRIPTION DIABLO CANYON COHANCllE PEAK WOLF CREEK CALLAWAY 14 07 [MedT__ Trt.secf- 3S b)i 43717,/-6 ' e. .m :a r. - 14-08hYotused. 'JA__s;d3B -- 8133"5?]?l-#1 r, r. .a - ;a I A
- 0' N EM..-_a-TA
"'" "Genese Aeta 1.00 is _acided linm&a.ftons on fusi-to storate. CTS 3/4 # o- N th tam;,y Poo9 -fd]l CTS I g no - m%
CTS %.'] # yCS
- I/ E S to 3.T.l& 3 14 10 The stat N Y a[ 3$0 M a h ble would be Yes No - CPSES does not Yes Yes A removed. have this specification in CTS 3/4.9 15 01 The requirement to empty the spent fuel exclusion zone area Yes - See No - CPSES does not No not in CTS No - not in CTS R prior to any spent fuel shipping cask handling operations Attachnent 21, page have this is relocated to I.icensee controlled documents. 27 telote.k l to specification in 4L FSAR CTS 3/4.9 Q set- M CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97
- -. = . . . .. - . . - - . .
NUREG 1431 SPECfFICATIONS THAT ARE NOT APPLICABLE . Soecification # Ijilt Comments 3.7.12- Emergency Core Cooling System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS) 3.7.14 Penetration Room Exhaust Air Cleanup System (PREACS) Bfon
& 3,9,% - 3 (
l \ l i i 1 g 'T., mae
- MARK UP OF WOG STS REV 1 (NUREG 1431) 2 5/15/97
Fuel Storage Pool Boron Ccncentration 3.7.16 Q 3.7.H,-3 { 3.7 PLANT SYSTEMS 3.7.16 Fuel Storage Pool Boron Concentration LCO 3.7.16 The fuel storage pool boron concentration shall be 2 [2S09] ppm. B-PS N RIG 5 APPLICABILITY: When fuel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 2 A. Fuel storage pool -
-------NOTE----- ----
baron concentration LCO 3.0.3 is not applicable, not within limit. --- ------ -- -- ---- A.1 Suspend movement of Immediately - fuel assemblies in the fuel storage , pool. AND A.2.1 Initiate action to Immediately restore fuel storage pool boron concentration to within limit. 93 A.2.2 Verify by Immediately a dministrative means g pg [ s[0 htorage ih j pool fuel gb verification has been p 1 performed since the non- last movement of fuel assemblies in the fuel storage pool.
~
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.__..e _ . __._.._. _ . . _._
i i i i Fuel Storage Pool Boron Concentration . 3.7.16
.m.;
. pv *
- q.g,. . SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY l 'SR 3.7.16.1 ' Verify the fuel storage pool boron 7 days concentration is within limit.
l
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l l i I t i l l 4 k ? . _
. ,,, M_ .,
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.T y .m- .,,.r- - p --e -re
y _ Fuel Storage Pool Boron Concentration B 3.7.16 4 3.7./(. - 3 B 3.7 PLANT SYSTEMS
'. 8 3.7.16 , Fuel Storage Pool Boron Concentration
'8ASES
/
f o BACKGROUND I t ax mu Denit/Ra ( R) (Refs 1a 2)] sig , E./ e sp t fu st ra ! p 1i di i et e ons whi h, or he di ided 'nto t o sep ate an rpos of c itica ty - [ %% i :p I
' ns'de at ns, are con ide ed sep rate ools. 7, M {' Re io 1 , wi h 36] sto age osit ons, 's des' ne to i
I P d* E g 3,#h0I' [a o od e n w'f el ith m imum enric ent ' [ .6 wt% L % r,
? -2 W!
, -sp nt uel .eg die s of the d'schar f 1 hu up. I gio 21 wi f2 70] toraa posi ion designed to accommodate fuel of various-initial enrichments gg which have accumulated minimum burnup,s within the acceptable ;, $9 domainaccordingtoFigure(3.7.17-1(,intheaccompanying Fuel assem ies not mieting the criteria of.
i i*b LCO. Figuref3.7.17-1 shall be stored in accordance with f. s paragraph 4.3.1. in Section 4.3, Fuel Storage. The water'in the fuel storage pool normally contains soluble baron, whic results in large subcriticality margins under actual operating conditions. However, the NRC 3 guidelines, based upon the accident condition in which all j soluble poison is assumed to have been lost,.specify.that the limiting k m of 0.95_be evaluated in the absence of NDE soluble boron. nence, tnetdesign W " --Sis based on the use of unborated water, .which iiiaintainsmted:M ( _ - in a suberitical condition during normal operationEwiThThe r v' The double contingency principle Q doge'[ WRi--Dfully discussed in loaded. ANSI N-16.1-1975 and the April 1978 NRC letter gh F Pes l re.c / (Ref. 3) allows credit for soluble boron under'other 4 j abnormal or accident conditions, since only a single e accident need be considered at one time. For-example, the 3 most_ severe accident scenario is associated with_the ! & ; m_; :: : : E- = g r - ' :: m:r. n. = accidental
- j misloading of>' fuel assem'~blv in c" xt; - ?P tis could j A j
muf$M T potentially increase the1;- ._m_. ._ of ercqn c Mo i itigate these postulated criticality related accidents, 7 j boron is dissolved in the pool water. Safe operation of the } TNg with no movement of assemblies may therefore be achieved y controlling the location of each assembly in accordance 1 ::s ', with LCO 3.7.17, " Spent Fuel Assembly Storage." Prior to p movement of an assembly, it is necessary to perform j
}{DR SR 3.7.16.1. *)
1
%e Gud skoese l l
-. s -
1
- l j (continued) i } l n = cc, m. .., s W
& W '
, - - - . . . - . _ = . _ . _ .
Q 3.7.16-3 l l INSERT B 3.7-81 1 i in the High Density Rack (HDR) design (Refs.1 and 2), each fuel pool storage rack location is designated as either Region 1, Region 2, Region 3, or empty (in the checkerboarding l configuration). Numerous configurations of region designation are possible. Criteria are ' l established for determining an acceptable configuration (Ref.1). The HDRs will store a maximum of 2363 fuel assemblies in the spent fuel pool and potentially an additional 279 fuel
' assemblies in the cask loading pool (with racks installed). Full-core offload capability will be maintained. The fuel storage pool consists of the spent fuel pool and the cask loading pool (with racks installed). Region 1 locations are designed to accommodate new fuel with a nominal maximum enrichment of 4.6 wt% U-235 with no integral fuel bumable absorber (IFBA);
or up to a nominal maximum enrichment of 5.0 wt% U-235 with 16 IFBA; or spent fuel j regardless of the discharge fuel bumup. Region 2 and 3 locations are ; l l ! l l ! l l I l i i
Fuel Storage' Pool' Boron Concentration i B 3.7.36 i BASES (c'ontinued) .s I 1 APPLICABLE t a iet on u ns do at esu in an cr%se n i SAFETY ANALYSES ea i f it r t t r to . am les f l _ hs c' et n ti s e e ss f ol 9 g e v% yin e e th ec eas'ng ate de it d he of Nis be' r pp ny o a = as em v tat o th r k Sbr e fool we erJJtcid_ents can be postulate _d that could increase the di owe, reactivityxWMr 9:rene " re rn g unacceptable with unborated water in the4 storage pool. Thus, for these Q accident occurrencest the presence of soluble boron in the storage cool f :x = : rntic:Ht. r ::= _ni xt The - postulated a'ccidents are basically of two tfpes. Sfuel T assemblicould be incorrectly transferred &= me'ei 2 to 3 W r 2"(e.g.,- an unirradiated fuel assembly or an- g 6 d c r M l' Insufficiently depleted fuel assembil). The second type of w;$. Kg,g d OM postulated accidents is associated with a fuel assembly which is droppet adjacent to the fully loaded C ::ter_ O f I 855* rack _._ S M., a Tc re: - -n : - 5
--- m vu . .J;in '" ._ "; =dAhe negative reactivity -- ---_-
_ % Jg
.8 # effTct of the solulle baron coupensates for the increased s
reactivity caused by either one of the two postulated pi accident scenarios. The accident analyses i rovided-in o the FSAR, @ 05.' j (Ref. f). i ~ ,yfspp. - The concentration of dissolved boron in_ th_e fuel storage 4 3 4.lb -pool satisfies Criterion 2 of mjgre Einff.
,Go crR 5'0.3McYh(td I w_ , _ -s LCO- The fuel storage pool boron concentration is required to be
;t [ p)m.# The specified concentration of dissolved 1 Agf oron in tie fuel storage pool preserves the assumptions !
used in the analyses of the potential critical accident ; scenarios as described in Reference C This concentration-of dissolved boron is the minimum required concentration for
; fu assembly _ storage and movement within the fuel storage I APPLICABILITY - This LC0 applies whenever fuel assemblies are stored in the L fuel storage pool, until a complete fuel storage oo verification has been performed follo@ wing the last 1b ' M Sy"he fuel storage pool.
Eg to66i5b p movement This LCO does of not fuelapply assemblies following in t the@he verification, since the verification would confirm that there are no misloaded O' ). fM cask fuel assemblies. With no further fuel assembly movements in Pa jh.t(wd d &D. (continued)
*-- 2
p Q 3.7.16-3 INSERT M Safety analyses assume a B-10 enrichment of 19.9 a/o (Ref.1). Administrative controls on the soluble boron concentration in the fuel storage pool ensure that there is equivalent B-10 concentration. L, '
-. . _ - _ - _ _ - - - . - - . - . . . _ . = _ - - - - . - . - . _ - - . - . . - . _ . . .
JCNCC L:CEW;;NC
- 163eui;6 - E 28i55 15 11 A .0C1-r A
[ [ - Fuel Storage ' col Boren Concentration B 3.7.16 i BASES , APPLICABILITY progress there is no potential, for@misloadedfuel , (continued) ass r a dropped fuel assgmbi r. { l
$$ i I l ?
. j- f ' -
~
,. ACTIONS ,
[ A.I. A.2.1. and A.2.2 j . f The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.
, When the concentration of boron in the fuel storage pool is
. e less than required, imediate action must be taken to
--4
- . : n preclude the occurrence of an accide;ht or to mitigate the
+- 6 .:.
~
l consequences of'an' accident in progr'ess. ' This is most l Z; O efficiently achieved by immediately suspending the movement ! seg. k. g. Y of fuel assemblies. The concentration of boron is restored l simultaneously with suspending moverent of fuel assemblies. l An acceptable alternative is to verify by administrative
,' means that the fuel storage pool ver ification has been I performed since the last movement of fuel assemblies in the fuel storage pool. However, prior to resuming movement of i
fuel assemblies, the concentration cf boron must be
, restored. This does not preclude acvement of a fuel assembly to a safe position.
If the LCD is not met while moving 60atfra't'e@ fuel o assemblies in NODE 5 or _6, tco 3.0.a woula not be applicable. If moving 0rfaO6tliD ftel assemblies while in NODE 1, 2, 3, or 4, the fuel movemerlt is independent of reactor operation. Therefore, inabi lity to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown. i SURVEILLANCE SR 3.7.16.1 REQUIREMENTS
- This SR verifies that the concentration of baron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time.
s m
- J (continued)
. . . . - ~ , - - . - , . . . - . - - . . . . - . - . . . . . - - . . . . . - . ~ - . - - - . _ . . -
- l. Fuel Storage Pool Boron Concentration - !
l- B 3.7.16 r
~
BASES (continued) $ , e .. . . k ~ ' h- REFERENCES 1. Callaway FSAR, Appendix 9.1A, "The *M Density Rack p R) Design Concept."
^
- 2. e cretio and EvMuati9fi for PfoposAd Cha e N ci sty per ign LicysesDP)r-39ffid DP -48 Zpi d j
_ o r5 ti n - - __.
- 3. Double contingency principle of ANSI N16.1-1975, as '!
.specified.in the April 14, 1978 NRC letter j (Section 1.2) and implied in the proposed revision to ,
Regulatory Guide'1.13 (Section 1.4, Appendix A). ' vu [1 I : m ; No.129 L le.A k iwca. l9,I999 h L ; 0AR Opp li m
- m. :
s '
- l i
r 1
.s'
~WOG STS'. B 3.7-84 Rev 1, 04/07/95 1
4 i
~ -- - . - -
, - . . . - , . , _ , _ , , . . . . , , , . i
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.8.10-01 DC APPLICABILITY: CP, WC, CA Q 3.8.10-01 CP Q 3.8.10-01 WC Q 3.8.10-01 CW REQUEST: CP(3.8.10-01) DOC 03-06-LS CW JFD 3.8-45 DC(3.8.10-01) ITS LCO 3.8.10 WC(3.8.10-01) STS LCO 3.8.10 CTS 3.8.3.2 Bases for ITS LCO 3.8.10, STS Bases markup page B 3.8-85 STS LCO 3.8.10 specifies, "The necessary portion of the AC, DC, and AC vital bus electrical power distribution subsystems ..." Corresponding ITS LCO 3.8.10 specifies, "The necessary portion of the Train A or Train B AC, DC, and AC vital bus electrical power distribution subsystems ..." JFD 3.8-45 states that the power distribution systems have been revised to retain the CTS requirement that one train shall be operable when shutdown. Corresponding CTS 3.8.3.2 states, "As a minimum, the following divisions of electrical buses shall be energized in the specified manner: " DOC 03-06-LS merely restates the proposed change. Comment: This LCO does not seem to be consistent with the Bases for LCO 3.8.10 or LCO 3.8.2. The LCO indicates that one train of AC, DC, or AC vital bus electrical power is required. However, the Bases for LCO 3.8.10 as well as 3.8.2 indicate that a second , train of AC, DC, or AC vital bus electrical power may be required. This inconsistency l needs to be addressed. Also, Action A.1 is an allowance to declare required features inoperable that is based on two trains of electrical power being required. Action A.1 is not appropriate for the LCO as written. JFD 3.8-45 and DOC 03-06-LS does not explain why the proposed difference is acceptable. Revise the submittal to provide the appropriate justification for the proposed difference. FLOG RESPONSE: (original) For DCPP see response to Comment Number O 3.8.10-04 DC. For CPSES, Callaway and WCGS, the ITS will be revised to reflect the CTS for the requirements associated with power supplies for supported systems during shutdown. FLOG RESPONSE: (revised) The markups associated with the CTS and ITS for this change are being restored to the version provided in the original license amendment request. However, the enhancement provided for DOC 03-06-LS-26 is retained; and, a new Bases modification is provided to clarify that only one train of Class 1E AC and DC power and distribution is required j to support equipment during shutdown conditions. ! l
- 1. The following two paragraph insert will be added in the Bases for 3.8.2,3.8.5,3.8.8 and l 3.8.10 (immediately preceding the last paragraph in the Applicable Safety Analyses section): <
l
1 In addition to the requirements established by the technical specifications, the plant staff must also manage shutdown tasks and electrical support to maintain ; risk at an acceptably low value. l As required by the technical specifications, one train of the required equipment during shutdown conditions is supported by one train of AC and DC power and distribution. The availability of additional equipment, both redundant equipment as required by the technical specifications and equipment not required by the specifications, contributes to risk reduction and this equipment should be i supported by reliable electrical power systems. Typically the Class 1E power sources and distribution systems of the unit are used to power this equipment because these power and distribution systems are available and reliable. When portions of the Class 1E power or distribution systems are not available (usually as a result of maintenance or modifications), other reliable power sources or l distribution are used to provide the needed electrical support. The plant staff assesses these alternate power sources and distribution systems to assure that the desired level of minimal risk is maintained (frequently referred to as maintaining a desired defense in depth). The level of detail involved in the assessment will vary with the significance of the equipment being supported. In some cases, prepared guidelines are used which include controls designed to manage risk and retain the desired defense in depth.
- 2. In the Bases for LCO 3.8.2 in the LCO section in the paragraph which starts out with words "The DG must be supporting . . ", make the following change. (Note that this paragraph had ,
been inserted by the FLOG during the original markup of the STS Bases). Replace the third j (and last) sentence of this paragraph with the following sentence. i i When the second AC electrical power distribution train (subsystem) is needed to support redundant required systems, equipment and components, the second train may be energized from any available source. The available source must be Class 1E or another reliable source. The available source must be capable of ' supplying sufficient AC electrical power such that the redundant components are capable of performing their specified safety function (s) (implicitly required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered.
- 3. In the Bases for LCO 3.8.10 in the LCO section in the second to the last paragraph (as added by the FLOG in the original submittal and which starts "The required DC electrical power . .)
add the following at the end of this paragraph: The available source must be Class 1E or another reliable source. The available source must be capable of supplying sufficient DC electrical power such that the redundant components are capable of performing their specified safety function (s) (implicitly required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered.
i
- 4. In the Bases for LCO 3.8.10 in the LCO section in the last paragraph (as added by the FLOG in the original submittal and which starts "The required AC vital bus electrical power . .") add the following at the end of this paragraph: ;
l The available source must be Class 1E or another reliable source. The available source must be capable of supplying sufficient AC electrical power such that the i redundant components are capable of performing their specified safety function (s) , (implicitly required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered. ;
- 5. Add the following paragraph to the end of the LCO portion of the Bases for 3.8.5. ;
The required DC electrical power distribution subsystem is supported by one train of DC electrical power system. When the second DC electrical power distribution train (subsystem) is needed to support redundant required systems, equipment and components, the second Train may be energized from any available source. The available source must be Class 1E or another reliable source. The available source must be capable of supplying sufficient DC electrical power such that the redundant components are capable of performing their specified safety function (s) (implicitly required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered.
- 6. Add the following paragraph to the end of the LCO portion of the Bases for 3.8.8.
l The required AC vital bus electrical power distribution subsystem is supported by one train of inverters. When the second (subsystem) of AC vital bus electrical power distribution is needed to support redundant required systems, equipment and components, the second train may be energized from any available source. The available source must be Class 1E or another reliable source. The available source must be capable of supplying sufficient AC electrical power such that the redundant components are capable of performing their specified safety function (s) (implicitly required by the definition of OPERABILITY). Otherwise, the supported components i must be declared inoperable and the appropriate conditions of the LCOs for the ! redundant components must be entered. ATTACHED PAGES:
- Encl. 2 Pages 3/4 8-8 and 3/4 8-12 Encl. 3A Pages 6 and 7 i Encl.3B Page 5 l Encl. 5A Pages 3.8-18,3.8-28,3.8-36, and 3.8-40 Encl. 5B Pages B 3.8-39, B 3.8-40, B 3.8-68, B 3.8-69, B 3.8-85, B 3.8-86, B 3.8-101, i
and B 3.8-102 Encl.6A Page 5 Encl. 6B Page 6 l,
b ELECTRICAL POWER SYSTEMS
)A.C. SOURCES '
SHUTDOWN 1 LIMITING CONDITION FOR OPERATION l 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be
]
OPERABLE: 1
- a. One circuit between the off ite transmission network and the Onsite Class IE Distribution Syst and ,,..
- b. One diesel generator-e...- m Y'NkNNhb -
N
) 7~
- 1) A d:y t:nk : nt:ining : =inimer vele : Of 510 4110n: ef'I:1,
-- 2 ) A feel : tor:g: :y:te cent:f 'ng a 64 e 19-* af
- 3. . ., . c. c. ....,
.., a Ol- Ol- A e
. n. , i. n n. . .. 3 .... . . .
- 2) A feel tr:n:fer ;;:p. r N .- Q 3 910-ol CW (bove \ s.w 4hvo , ;
APPLICABILITY: MODES 5 and 6.
"^b ACTION:
Itdert] g With le:: th: th: :bev =inimer required A.C. :10 tri::1 ;:,wer :: rt: 61-59-L5-2o OPE"ASLE, positive reactivityimediately changes, movement suspend all operations of irradiated fuel,"cr cr:n: involving C01E ALT Ope tien o!-44-LG,. eith lead: Over the :p:nt ft: 1 p::1. In :.dditi:n, wher, in ".00; ; ith ; re::t:r :0:1:nt-10:p: not filled, er in "00E S with the r:ter level le::jeth:n o[.45-M - l 22 feet abeve the re :ter ve:: 1 '!:nge, imediately initiate corrective action to restore the re' quired sources to OPERABLE status as soon as possi ei ' CTAse.rt h [MM SURVEILLANCE REOUTREMENTS 01-43 N\ l 1 4.8.1. The :b;ve requir:d A.C. 10:tri::1 ?:r:r :curce: :h:1' be i d 20nstr:ted OPEPABLE by the perf:-.:nce Of ::ch of the require::nt: e f- ol-47-L5 4 Specificstiens 4.0.1.1.1, 4.3.1.1.2 ( :<cpt fer Speci fic:tien ' .S.1.1.2 .5),
- nd
'.S.1.1.29 2), '), S) :nd 10) for th- LOCA : quer.cer pertier of LSELS- 1
.IN< f4.3 IMettj CALLAWAY - UNIT 1 3/4 0-8 Amendment No. 8&rMO,112
i l
) INSERT 1 j Ol-46-R l one required offsite circuit inoperable, enter applicable ACTIONS of LCO 3.8.3.2 and Oi-MS-2o declare affected required features with no offi.ite power available inoperable, or l
l b uth h.fhr g 3.g,j o_o t ([ l INSERT 2 With one requireil diesel generator inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement ofirradiated V /l[4hu GQ , fuel, immediately initiate corrective action to restore the required diesel generator to o15g.L5 2o OPERABLE status as soon as possible. ! Enter applicable ACTIONS ofLCO 3.8.3.2 if one train is de-energized as a result of g, g having the required offsite circuit m
' operable.
INSERT 3 l NOTE The following Surveillance Requirements are not required to be performed: i 4.8.1.1.2a.5),4.8.1.1.2g.1),2),6),7), and 10) for the shutdown sequencer portion of ol-4'l-LS-+ LSELS only, and 4.8.1.1.2g.11). . I I I INSERT 4 For AC sources required to be OPERABLE, the following requirements are applicable: l 4.8.1.1.1 4.8.1.1.2f l 4.8.1.1.2a.1 4.8.1.1.2g.1 0 (-4MS-+ l 4.8.1.1.2a.3 4.8.1.1.2g.2 { 4.8.1.1.2a.4 4.8.1.1.2g.6 ; 4.8.1.1.2a.5 4.8.1.1.'2g.7 4.8.1.1.2b 4.8.1.1.2g.10 (shutdown sequencer only) ,
. 4.8.1.1.2c (shutdown sequencer only) ' 4.8.1.1.2g.11 !
I 5 i
r - b[b_ d 3.8.10-01 CW ELECTRICAL POWER SYSTEMS ' 4ke. Treo.n A e r Train 5 D C c l< c. a.1 D_.C. SOURCES 5.4sys k sb.tt k orca.Astt h s g.: r*< 1 , 64 re. wivd
%= by DCt.co cle7.S.3.1 c4Sce.] ?*t Poi M didv'abdth ee olde!I A4 swb5ydams SHUTDOWN
[ u s - sw.,w .wn "
" oM.h.
LIMITING CONDITION FOR O PERATIun j l 3.8.2.2 As a minimum,V::: Of th: f:ll:ef ; D.C. :10:tri::1 ::er::: :h:11 he 7,g3.A o 1PEftABtt:
- . 125 ":11 D.C. h:::: HK01 :ad NK02 :::rgi::d frea. S:tterie: NK11 :nd NX 2 :nd :::::i:ted hil-C:p::ity Ch:rg:r: NX21, NX22 :r in:t:11:d
' ref t; Ch: ; r M"25 perered 'r:: M201, er 02.-ot-LG 5_ ? ?5-Yelt D.C. B'arter w 02 and ""0' erergi:ed 're- B:tterie: ""12 : d-NX 4 :nd e:;;;iated N11 C:;::ity Ch:r; r: ""22, "K?4 er %:t:11:d _
uing Ch:r;;r Nfl5 p: :r:d fr : M:0'. i APPLICABILITY: " MODES 5 and 6. . immd:.hh L.ctus =W.cW regu. d ir ACT10N: k hveta) 6 pu Ma.or, i i k'ith the required D.C. electrical source inoperable,kimmediately suspend all og_tus.zo operations involving CORE ALTERATIONS, positive reactivity changes or movement , ) ofelectrical irradiated fuel; source initiate corrective to OPERABLE action status as soon to restore the required D.C. as possible. 2 SURVEILLANCE REOUIREMENTS . w 4.8.2.2 The above required D.C. electrical source s 11 be demonstrated . DPERABLE in accordance with Specification 4.8.2. .
- ot_ts.g 4 l
* % fn,tt.J.g sns <
ut <=piva b 6 p6 mea : 4.8. 2.i c. 4, 4.t.1.t 6, 4.e.t.: a , ma 4.s.2.1 G . CALLAWAY - UNIT 1 3/4 8-12 Amendment No. 99
l l p i CHANGE NUMBER NSBC DESCRIPTION i 01 33 LS-15 Not applicable to Callaway Plant. See Conversion > Comparison Table (Enclosure 3B). ' 01 34 Not used. 01 35 "M Not applicable to Callaway Plant. See Conversion g 3.~8. )-05 Cd
. Qomparison Table (Enclosure 3B).
01 36 . The pow r fa tor ower imi of .8 ld de ted i
- f. the - r fac r 11 ts i be n .8 0 . Ti c nge i con ist t wi G-1 31. -y 01 37 LG The specific SRs for sampling new diesel fuel oil (DFO) !
prior to addition to the storage tanks and for sampling ; the stored DF0 would be moved to the Diesel Fuel Oil 1 Testing Program.-which would.be referenced by ITS SR 3.8.3.3 and described in ITS 5.5.13. The Diesel Fuel . 011 -Testing Program will function similar to other
~
l programs currently described in the Administrative Controls section of the CTS. SR 3.8.3.3 would require i that DF0~ properties.of new and stored DF0 are tested in i
~3, '4 i accordance with>and maintained within the limits of. the l Diesel Fuel 011 Testing Program. In addition, the
~
j SR 3.8.3.3 Bases will contain additiGnal descriptive material regarding the DFO. test % requirements of the current SRs.~ These' changes are consistent with NUREG-1431. --
.v..,
01-38 - LG "Thehuel oil storage tank cleaning SR would be moved to
..thelFSAR]. .This change is consistent with NUREG 1431.
3-w 01-39 LG Not applicable to Callaway Plant. See Conversion l Comparison Table (Enclosure 38). 01-40 LS 25 The rehuirements for staggered DG testing would be deleted l and replaced with a requirement to test each DG every 31 days. Deleting staggered DG testing is consistent with NUREG 1431 and results in no change in the frequency of i testing for each DG. This change is acceptable because l studies have shown that staggered testing has negligible impact on equipment reliability. It is not as safety significant as originally thought, it.is difficult to 1 implement operationally, and may increase the chance of
. human error ~ test performance.
[ Q 3,g,[0-bl CW ; 01 41' M ur es - S t L ld an cif t eq r AC s rc st 9. or a e j
, . DESCRIPTION OF_ CHANGES TO CURRENT TS 6 mow LO-N 5/15/97 ,
L- I ._ ~ ~ 1.1 - _____,_..._.___m._ __.;_ .
~ __ . _ _ __ _ _ _ _ _ _ . . _ . . __
- C "
- CHANGE NUMBER HSHC DESCRIPTION
[of ng, t is an dditio requir d AC stribu on bus es. T is res iction onsist t wit the fchan imp ' cit a umptio for ratio durin hut j c itio . Thi cha is con stent th N G 14 . 1 See al CN 0 LS .) _ 01 42 M Not applicable to Callaway Plant. Wmox See UMhW() Conversion 3 B. Hod (N' i Comparison Table (Enclosure 3B). 01 43 H The AC Sources - Shutdown Action statement would be
- revised to specify that the offsite circuit must be capable of supplying power to the required onsite buses.
l This assures the single operable circuit is perforuing a vital function. Explicitly requiring entry into the applicable'actuons of LCO [3.8.3.2] clarifies the l appropriate action-to take if the onsite bution l train becomes de energized. This consistent w h l NUREG 1431. hMe , Q 3 81-02 C%- l 01-44 LG The AC Sources v-Shutdown actio1 statement , reference to ; l crane operations with loads ov the spent fuel pool ld s.hThisis i _1 . be moved to licensee controlled (Q,9 consistent with ElREG 1431 and is example vi moving details not required for operational out of . 01-45 M The AC Sources - Shutdown Action statement requirement to initiate corrective action is expanded to match the applicability of the CTS (Mode 5 and 6). This will replace the' current requirement to *iumediately initiate corrective.. action"gwhich is applicable only if RCS loops not full in Mode 5 and less than 23 feet of water above the reactor vessel flange in Mode 6. This more restrictive requirement continues to ensure that qualified power is provided to systems required operable in these modes.' This change is consistent with NUREG 1431. 01 46 A A note would be added to the AC Sources - Shutdown action statement to enter the Distribution Systems LC0 if one required electrical distribution subsystem has no electrical power. The change is already implicitly a part of the CTS requiring one offsite circuit and the onsite
- distribution system required to support Onsite Power Distribution - Shutdown LCO. The added note clarifies
) that the need is to support necessary equipment. This
, change does not add or remove any technical requirements 4 and is therefore achinistrative in nature. This change is consistent with NUREG 1431.
DESCRIPTION OF CHANGES TO CURRENT TS 7 5/15/97
i 4
. j .
, CONVERSION COPFARISON TABLE CURRENT TS 3/4.8 Page 5 of 11-TECH SPEC CHANGE ' APPLICABILITY lIi ,
NUPBER DESCRIPTION DIABLO CANYON CONANCHE PEAK- WOLF CREEK CALUNAY u j 01-39 Fuel oil transfer piping leak test, would be sowed to Yes Yes: moved to TRM.- No; not in CTS. No: not in CTS. LG licensee-controlled documents implementing the Inservice ' '
.y!
i Inspection Program. i i
' li -
p 01*40 The requirement for staggered DG testing would be deleted Yes Yes Yes Yes LS-25 and laced with requirement to test each DG every E Q3B.lo-olCW) . 01-41 AC Sources Shutdown LCO would be changed to spectfy that Yes Joe'$W +g Yo( : , ;- h S-[.Yes'NOj i M,.] 4pS- yeS M the required AC sources must be powering, or capable of g_g_ 7eS gg yeS i powering the required AC distribution bussesi =i - - -
! 01 CTS 3.8.1.2. CTS 3.8.2.2 and CTS 3.8.3.2 applicsbility .
Yes I ho: maintain CTS. No: maintain CTS. No: not in CTS. M would be revised to add "during sovement of;1rradiated fuel : i assemblies'- 't' l
\
01-43 ACSources-Shutdownactionstatementwouldberevisedte! Yet .! !' Yes Yes Yes M specify that the offsite circuit must be captble of supplyingpowertothercquiredonsitebuset;and.- ! . therefore to all equipment required to be operable in the j , applicable modes. .q 01-44 Reference to crane operations with loads over the spent Yes: moved to plant Yet: soved to the Yes Yes: moved to the -I LG fuel pool would be moved from the LCO action to licensee- procedures. Bases. FSAR. controlled documents. - - 01-45 The requirement to initiate corrective action is expanded Yes Yes
~
Yes Yes M to match the applicability of the CTS. 01 46 A note would be added to the AC Sources - Shutdown action Yes Yes Yes Yes A statement to enter the Distribution Systems LCO if one required train has no electrical power. 2 01-47 The SRs required for AC sources operability in Modes 5 Yes Yes Yes Yes LS-4 and 6 would be nevised to include only those SRs which are , applicable. The Note listing exceptions to SR required for ' j' Modes 5 and 6 in current TS 4.8.1.2 would be revised to include additional SRs.
! CONVERSION CONPARISON TABLES CURRENT TS 5/15/97
l - , , ,
~ -~
~ '
AC Sources Shutdown . 3.8.2 ( y
- 3.8 ELECTRICAL POWER SYSTEMS n
3.8.2 E Sources-Shutdown
/
= *
'OIk
~
LCO 3.8.2 following E electrical" power sources shall be OPERABLE:
- a. One qualified circuit between the offsite transeission network and the onsite Class IE E electrical power distribution subsystenfeb R3.8 45E {
required by LCO 3.8.10. " Distribution Systems-Shutdown"; and , l
- b. One diesel generator (DG) capable of supplying one train of the onsite Class 1E AC el distribution subsystentsF n 3'8 45 s required by LCO 3.8. ;ud 3
> f 8- 07 f 3.3-99_\
APPLICA81LITI: MODES 5.and 6 . .. . . ,
. %-i 6.;; _ 3 n Ri.diid fal .:-:Miin. t&3.8 29 4 i r " j J
~ '
" ACTIONS , -
.c,, , COISITION a REQUIRED ACTION COWLETION TILE s .-
. l A. One required offsite ----
- --- NOTE ---- ----- -- !
circuit inoperable. ~ Enter applicable Conditions and I
~~
Required Actions of LCO 3.8.10 iitti ~ene required train a:3.8-45d de-energized as a result of Condition A. A.1 Declare affected required Inmediately feature (s) with no offsite power available inoperable. El A.2.1 Suspend CORE Imeediately ALTERATIONS. . i ale !
\ (continued) !
i
. a.s, c. % ~ Abn 7.v h".6 of ou L.ca Shd de r x1 Emerp.,cy g 3,j_g l Q9 L. b 4.g .,* n ca.r (Lsels) anoci&L wh W n
. . . os mh At el.Jma p w 24ru.oh.n 4rdn . glir 1 i l
MARKUP OF WDG STS REV 1 (NUREG 1431) 3.8 18 5/15/97
.e
. sa :, p DC Sources Shutdown .
.- 3.8.5 ft )
\_ l I
3.8 ELECTRICAL POWER SYSTDtS 3.8.5 DC Sources - Shutdown
~
LCO 3.8.5 ' - DC electrical subsystem shall be c3.8 45 OPERABLE to support- the DC el .rical power distribution subsystestsi- required by LCO 3.8.10, "Distritution Systems - i 4 ( Shutdown *" {Q 3 8.1D-O l CW APPLICA8ILITY: MODES 5 and 6-irir.; ;^._c.; ef ir. di;ted fel ;;;2iin. e3.8 29 a
~
ACTIONS ~i -
- ~~
COWITION ~ REQUIRED ACTION COWLETION TIE A. On er de pequired DC A.1 rl- - Declare affected Immediately + e ED.% s 1 electrical power required feature (s) 43.8459
"' 1
, subsystees inoperable. inoperable. -
~~
GL A.2.li Suspend _. CORE Immediately ALTERATIONS. M A.2.2 Suspend movement of Immediately irradiated fuel assemblies. I 1 M A.2.3 Initiate action to Ismediately ! suspend operations 1 involving positive reactivity additions. 1
. -l
.- l eo . .
A% ' l (continued) i 1 HARKUP OF WOG STS REV 1 (NUREG 1431) 3.8 28 5/15/97
<.m.._ . _ - _ . . _ . _ ___, _ . . . . . _ . . . _ . . . fM 4 * "'"9 ..y'aJ T. 2
- *m s a
Iny;rters Shutdown . 3.8.8 l
'4 l
3.8 ELECTRICAL POWER SYSTEMS 3.9.lb -D l CW t l 3.8.8 : Inverters -Shutdown
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i l l LCO 3.8.8 mverters shall be OPERABLE to support M W3.8 45~3 : M the onsite Class 1E AC vital bus electrical power l distribution subsystents)- required by LCO 3.8.10. " Distribution ;
- Systems -Shutdown."
, i l
! APPLICABILITY: H0 DES 5 and 6 l - .. , . - ... .. ... -.. .. .-. . . - ... . m3.8 29W l~ .
- n. ALTIONS
, , " T. ~Z~' 1 ' ~ COWITION I REQUIRED ACTION COMPLETION TIE J-s
~ ~~
_ _ . . i A. One or more - A.1 Declare affected required Immediately <1BN[~ l inverters inoperable, feature (s) inoperable. (O':s
' ;.cw . * ..
A.2.1 Suspend CORE Immediately . - - - -- l l-l A.2.2 Susgnd movener.t of Immediately l l 1; radiated fuel . assemblies. M. A.2.3 Initiate action to Innediately
* .. , suspend operations :
involving positive 1 l reactivity additions. _
. E
.p.p;, (continued)-
L~~,q, . l , ,. y . I HARKUP OF RG STS REV 1 (NUREG 1431) 3.8 36 5/15/97
, 7_.. _ - - - - .
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DistributionSystemso Shutdown . o 3.8.10 je .. 3 . g 3.8 ELECTRICAL POWER SYSTEMS Q 3 S.l0-01 CW ( 3.8.10 Distribution Systems; Shutdown
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l LCO 3.8.10 The necessary portion of AC, DC, and AC vital c 3.8-45 <- bus electrical power distribution subsystems shall be OPERABLE to i support equipment required to be OPERABLE. k i APPLICABILITY: MODES 5 and 6 irir.; 2._./. cf iris.di.ted fal n;dlic;. 13.8 291:4 ACTIONS C000ITION - C ~ ' REQUIRED ACTION CONPLETION TILE A. One or more required AC, A.1 ~ Declare associated supported Inmediately - DC, or AC vital bus- . .; required feature (s) electrical power .. inoperable.
; distribution subsystems.
inoperable. {R A.2.1 Suspend CORE Immediately
~
_.- ALTERATIONS.
~
AIE
]
A.P.2 Suspend movement of Immediately ) irradiated fuel ! assemblies.
. ggg i A.2.3 Initiate action to Immediately suspend operations involving positive reactivity additions. i blQ .l (continued) !
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i ff f.s,
.i MARKUP 0F WOG STS REV 1 (NUREG 1431) 3.8 40 5/15/97-
m AC Sources - Shutdown . B 3.8.2
, . .X:s BASES APPLICABLE testing and maintenance activities must be conducted provided an SAFETY ANALYSES acceptable level of risk is not exceeded. During MODES 5 and 6, (continued) performance of a significant number of required testing and maintenance activities is also required. In H0 DES 5 and 6, the activities are generally planned and administrative 1y controlled.
Relaxations from H0DE 1, 2, 3, and 4 LC0 requirements are " acceptable during shutdown modes based on:
- a. The fact th'at time in an outage is limited. This is a risk prudent goal as well as a utility economic consideration.
- b. Requiring appropriate compensatory measures for certain
-~
conditions. These may include assinistrative controls. - a reliance on systems that do not necessarily meet typical '
- =
~ .-_ -~
design requirements applied to systems credited in - -- operating MODE analyses, or both. .
- c. Prudent utility consideration of the risk associated-.with=
l y# &. y.. multiple activities that cu;ild affect multiple systems.
- 5. e
"~
l 70 '
- d. Maintaining, to the extent practical, the ability to l
- perform required functions (even if not meeting NODE 1. 2 - ~
3, and 4 OPERABILITY requirements).with systems assumed to a: function.during an event. ;9 i; - - In the event of an accident during shutdown, this LC0 ensures .the
- _ __ capability to support systems necessaryD avoid immediate[jr jr
'fr difficulty, assuming either a loss of all.1offsite power or.a lossi :=
Y, 2 of all onsite diesel generator (DG) poseiU~ 3ih5~ ~ 9
}v\tectly-p
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- 1. ,
~
m -- The AC sources satisfy Criterion 3 of " -" " ' 't---t i
...o- :. .. 1 j g3 9.lo-01 CWk
/- N LCO One offsite circuit capable of supplying the onsite Class 1E power distribution subsystests-)- of LCO 3.8.10. " Distribution Systems-Shutdown.* ensures that en M required loads 1 are powered from offsite power. An OPERABLE DG, associated with j the distribution systes train required to be OPERABLE by j
-- LCO 3.8.10, ensures a diverse power source is available to i 1. I provide electrical power support, assuming a loss of the offsite l
% circuit. Together, OPERABILITY of the required offsite-circuit l M I i
(continued) MARK UP OF NUREG 1431 BASES B 3.8 39 5/15/97
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ws. - s , a < mNe$ ,, u- .,w. . . . - .
.e-degh@-da,g-W-- i *M r 1= - , W =w.7 **+1-d=' ---m if-- -'-i-ue to4 4=r4.--a 1 t* + - - * -
INSERT FOR Q 3.8.10-01 CW Enclosure SB, Page B 3.8-39 in addition to the requirements established by the technical specifications, the plant staff must also manage shutdown tasks and electrical support to maintain risk at an acceptably low value. As required by the technical specifications, one train of the required equipment during shutdown conditions is supported by one train of AC and DC power and distribution. The availability of additional equipment, both redundant equipment as required by the technical specifications and equipment not required by the specifications, contributes to risk reduction and this equipment should be supported by reliable electrical power systems. Typically the Class 1E power sources and distribution systems of the unit are used to power equipment because these power and distribution systems are available and reliable. When portions of the Class 1E power distribution systems are not available (usually as a result of maintenance or modifications), other reliable power sources or distribution are used to provide the needed electrical support. The plant staff assesses these altemate power sources and distribution systems to assure that the desired level of minimal risk is maintained (frequently referred to as maintaining a desired defense in depth). The level of detail involved in the assessment will vary with the significance of the equipment being supported. In some cases, prepared guidelines are used which include controls designed to manage risk and retain the desired defense in depth. I 1 i
l i AC Sources - Shutdown . B 3.8.2 BASES
- LCO and DG ensures the availability of sufficient AC sources to j (continued) operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents).
\ g3ggg,og cw \
i $ The qualified offsite circuit must be capable of maintaining i , rated frequency and voltage, and accepting required loads during j an accident, while connected to the Engineered Safety Feature
- (ESF) bus (es). Qualified offsite circuits are those that are i described in the FSAR and are part of the licensing basis for the j unit. ' i
)
l cffsite circ.it n car.;ists of car;g.sr.n Tr.r.srerr r c. t.ich is
- m-aud fr = e,rd =
- . ..- a =d =._ t br= =r = :
i g-;r%;; tt Ci trere'erar L'=, Jid,17. turr., p ;re tra j n ZI b.- tt.ra f,it Ter;.s.1 f;;dir bra'ar. Oc x ar.d eff-it;
- ;irait ;.cai;t; ef tr Stertai; Treraferar, d.id i; recally
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1 pa;rir.ii tr.; CI trerafeiar, 2.13, ir. tu r., pwer., tte M CT j be; tt.rx @ it, seral fad,.c br;d;r. l The DG aust be capable of starting, accehrating to rated speed j and voltage, and connecting to its respective ~ ESF bus on j detection of bus undervoltage. This sequern must be accomplished within S61 g seconas. The DG aust be capable of accepting required loads within the asstaned loading sequence ; intervals, and continue to operate until offsite power can be j restored to the ESF buses. These capabilities are required to be
- net from a variety of initial conditions such as DG in standby with the engine hot and DG in standby at ambient conditions. !
l 4 i Proper sequencing of loads, including tripping of nonessential l
.c;& loads, is a required function for DG OPERABILITY. "c< l i
) i (continued) i < MARK UP OF NUREG 1431 BASES B 3.8 40 5/15/97 l
l t INSERT FOR Q 3.8.10-01 CW. . Enclosure 58, pgg,9 3 g 0 When the second AC electrical power distribution train (subsystem) is needed to support redundant required systems, equipment and components, the second train may be energized from any available source. The available source must be Class 1E or another reliable source. ; The available source must be capable of supplying sufficient AC electrical power such that the , l redundant components are capable of performing their specified safety function (s) (implicitly ! l required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered. l l l l i 1 l l r i
4-j i DC Sources - Shutdown . i B 3.8.5 1 4 I .ye l m, ,'3 =e. -.' [. I s E i . "+,a ) , ,',' i . 'I e jgj c...~,y . s .'.,.
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G 3 9.(0-01CW ff%ethD t - I DC sources satisfy Criterion 3 of the ",0 Mih St.t_..t E
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g'ggg g i 7 Q 3 9,lO-ol CW) E (continued)
~ -
MARK UP OF NUREG 1431 BASES B 3.8 68 5/15/97
t INSERT FOR Q 3.8.10-01 CW - Enclosure 5B, Page B 3.8-68 ;
- In addition to the requirements established by the technical specifications, the plant staff must also manage shutdown tasks and electrical support to maintain risk at an acceptably low value.
As required by the technical specifications, one train of the required equipment during shutdown ! conditions is supported by one train of AC and DC power and distribution. The availability of : additional equipment, both redundant equipment as required by the technical specifications and , equipment not required by the specifications, contributes to risk reduction and this equipment should be supported by reliable electrical power systems. Typically the Class 1E power sources and distribution systems of the unit are used to power equipment because these power and
- distribution systems are available and reliable. When portions of the Class 1E power distribution systems are not available (usually as a result of maintenance or modifications),
other reliable power sources or distribution are used to provide the needed electrical support. The plant staff assesses these attemate power sources and distribution systems to assure that'
- the desired level of minimal risk is maintained (frequently referred to as maintaining a desired defense in depth). The level of detail involved in the assessment will vary with the significance of the equipment being supported. In some cases, prepared guidelines are used which include controls designed to manage risk and retain the desired defense in depth.
i 1 l I I A l c-
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r l l ! pm L,a-fhru) DC Sources - Shutdown B 3.8.5 (~ ~ J Q 3.t3.(0-ol cW CfJG u d n ! 8ASeS l f 1 LCO
;' t.' ;r; ;gir;d te be OPERABLE to __suppnrt
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q a 4g I 4 34. I E U b5 E. U 5 94. WI b 5.to M U eF b3 5.w & 5 5. eFg er D use. . hqw 5 . t.M ans e L.. e en on,_A_,L.A,__ e. et.22_._ e.
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....;.c..,...n., . ...o., ,.....,. ..c...,n. This ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to sitigate the consequences of l postulated events during shutdown (e.g., fuel handling a I i g 3,g,g o_o 1 CW
- snsa l-,qidents).
( l 3 333 3 -
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mammmmmun m masensmem ummmmmemn EmEEmImme memquume mammmmmun munuman-m usumuseos m p======nu m e APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 5 and 6. .c. " rir.; ;.;r.; .,; ;f ir. .di;ted 7.;l ;;;z-'11;;. - provide assurance that:
- a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core, l I
- b. Required features needed to mitigate a fuel handling accident are available;
- c. Required features necessary to sitigate the effects of events that can lead to core damage during shutdown are available: and O.y -
(continued) MARK UP OF NUREG 1431 BASES B 3.8 69 5/15/97
,. . ~- . - ---.- . . - _ . . -_ . -- . --
l l lNSERT FOR Q 3.8.10-01 CW Enclosure SB, Page B 3.8-69 The required DC electrical power distribution subsystem is supported by one train of DC electrical power system. When the second DC electrical power distribution train (subsystem) is needed to support redundant required systems, equipment and t components, the second Train may be energized from any available source. The ! available source must be Class 1E or another reliable source. The available source l must be capable of supplying sufficient DC electrical power such that the redundant ! components are capable of performing their specified safety function (s) (implicitly - required by the definition of OPERABILITY). Otherwise, the supported components ' must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered. i l i j l l i 1 1 l i j ; i I l l l l
4 k + Inverters - Shutdown . i B 3.8.8 1 j , . - l E l APPL.IN E .e i. . . i s t ~ 3 -i w. e : > -. . . .
. . . -. .' .t.. .. i.- -. .
I SAFETY ANALYSES m i w a .. ,x i. r. . . m.. . . i c, .
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j , Q 3.B. Lo-ol cw , 1_..q. 1A-_2 2n1_A .. _ _x _a tL-The 1nyerters _- . r__. . . . . . . , . . . . . . . _ .. r. . . . . ..
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.._..e N i St.t ..; tir m - M w a m m i.
(continued) MARK UP OF NUREG 1431 BAS'S B 3.8 85 5/15/97
INSERT FOR Q 3.8.10-01 CW Enclosure SB, Page B 3.8-85 I In addition to the requirements established by the technical specifications, the plant staff must also manage shutdown tasks and electrical support to maintain risk at an acceptably low value. As required by the technical specifications, one train of the required equipment during shutdown conditions is supported by one train of AC and DC power and distribution. The availability of additional equipment, both redundant equipment as required by the technical specifications and equipment not required by the specifications, contributes to risk reduction and this equipment should be supported by reliable electrical power systems. Typically the Class 1E power sources and distribution systems of the unit are used to power equipmant because these power and distribution systems are available and reliable. When portions of the Class 1E power distribution systems are not available (usually as a result of maintenance or m>difications), other reliable power sources or distribution are used to provide the needed electrical support. The plant staff assesses these alternate power sources and distribution systems to assure that the desired level of minimal risk is maintained (frequently referred to as maintaining a desired defense in depth). The level of detail involved in the assessment will vary with the significance of the equipment being supported. In some cases, prepared guidelines are used which include controls designed to manage risk and retain the desired defense in depth.
M ove L. da. Jhr v Inverters - Shutdown . B 3.8.8 _. Q3.9.10-0lCvJl
- BASES (continued -
W m . . .; . . .. ....,c.. - 1 .' s . 's..
'6 ,s. i,.r. . .. . , . . . .i..
4.'Io, ..~.:y53 ...je'.j a i ...,_. ; , , , , , t.r n,r.~, .,..,c..
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The inverters ensure the availability of electrical power for the instrtamentation for systems required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated DBA. The battery pc rM inverters provide uninterruptible supply of AC electrical power to the AC vital buses even if the 4.16 kV safety buses are de energized. OPERABILITY of the inverters requires that the AC vital bus be pc.c.rM by the inverter. This ensures the availability of < sufficient inverter power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events awho shutdown (e.g., fuel handling accidents). i
-i Q%b.lO-olCVJ}
hNut- >> J gangs ,,m . C . m ammmes
- m t . .. .. ,.. ; . . . . .. ...
.. " .. u .o .7.. ., ,, . . . . o m
- M M M M
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, . ...,n.
APPLICABILITY The inverters required to be OPERABLE in MODES 5 and 6 .c.: 0.717.;
;;.;;. :.r.t ;f irc.di;t;d f.;; ;;::-t1;;; provide assurance that:
- a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core:
- b. Systems needed to mitigate a fuel handling accident are available;
- c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available: and
- d. Instrissentation and control capability is available for monitoring and maintaining the unit in a cold shutdown 1 condition or refueling condition.
4 (continued) MARK UP OF NUREG 1431 BASES B 3.8 86 5/15/97 i
i [: i l ' l l < INSERT FOR Q 3.8.10-01 CW , e Enclosure 5B, Page B 3.8-86 ! l l The required AC vital bus electrical power distribution subsystem is supported by one ! l train of inverters. When the second (subsystem) of AC vital bus electrical power ! distribution is needed to support redundant required systems, equipment and ' components, the second train may be energized from any available source. The available source must be Class 1E or anotha reliable source. The available source must be capable of supplying sufficient AC electrical power such that the redundant ! components are capable of performing their specified safety function (s) (implicitly I required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered. i i l i 1 l { i f k I i
4 Distribution Systems Shutdown . B 3.8.10 t J_ j l - BASES APPLICABLE 1 WF. . tin % t407. .- h Dr V J. i.i u u
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i SAFE 1Y ANALYSES m , , ": m k m B., w r t e w. .i. m o ! . v t ' .. m av-3 w s
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EDn3MEEEMB : 4 i FI M tidf0B M 5tRI E M M EE r SnwQ _ The AC and DC electrical power distribution systems satisfy Q *5.9. lo-o \ Cu ) ; Criterion 3 of the = Mky ';tet ./. twad!wM.pmup. . I
- )
1 (continued) . MARK UP OF NUREG 1431 BASES B 3.8 101 5/15/97 l l 1- ' i
c.- - . . . - . _ ~ -.- . . - - - .-. - . . . - - - . - . - - i INSERT FOR Q 3.8.10-01 CW Enclosure 58, Page B 3.8-101
' In addition to the requirements established by the technical specifications, the plant staff must
- also manage shutdown tasks and electrical support to maintain risk at an acceptably low value.
As required by the technical specifications, one train of the required equipment during shutdown
- conditions is supported by one train of AC and DC power and distribution. The availability of additional equipment, both redundant equipment as required by the technical specifications and i equipment not required by the specifications, contributes to risk reduction and this equipment should be supported by reliable electrical power systems. Typically the Class 1E power sources
- and distribution systems of the unit are used to power equipment because these power and distribution systems are available and reliable. . When portions of the Class 1E power distribution systems are not available (usually as a result of maintenance or modifications),
l other reliable power sources or distribution are used to provide the needed electrical support. l The plant staff assesses these attemate power sources and distribution systems to assure that the desired level of minimal risk is maintained (frequently referred to as maintaining a desired defense in depth). The level of detail involved in the assessment will vary with the significance l of the equipment being supported in some cases, prepared guidelines are used which include controls designed to manage risk and retain the desired defense in depth. l i i_ ). k
j~ Distribution Systems Shutdown . Q 3.9.10-0 ( CW l BASES (continued) _ l l LCO .Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant I condition. Implicit in those requirements is the required i OPERABILITY of necessary support required features. This LCO
, explicitly requires energization of the portions of .-' +. ! the electrical distribution system necessary to support l OPERABILITY of .
required systems, equipment, and i components all specifically addressed in each LCO and i implicitly required via the definition of OPERABILITY. Maintaining these portions of the' distribution system energized l ensures the availability of sufficient power to operate the unit I in a safe manner to mitigate the consequences of postulated
. events during shutdown (e.g.. fuel handling accidents).
J ' *. . .
- ,, . . +
w :.. .
,,p.. ~
@g APPLICABILITY The AC and DC electrical power distribution subsystems required -
to be OPERABLE in H0 DES 5 and C, erd ducir.;; =nant of irrediet;d fal ;;x;ilic;. provide assurance that:
- a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core;
- b. Systems needed to mitigate a fuel handling accident are
;;}!"Ey) available:
(continued) MARK UP 0F NUREG 1431 BASES B 3.8 102 5/15/97
I INSERT FOR Q 3.8.10-01 CW Enclosure SB, Page B 3.8-102 Insert i The available source must be Class 1E or another reliable source. The available source must be capable of supplying sufficient DC electrical power such that the redundant components are ! capable of performing their specified safety function (s) (implicitly required by the definition of l OPERABILITY). Otherwise the supported components must be declared inoperable and the
]
appropriate conditions of the LCOs for the redundant components must be entered. i
\ 8, Page B 3.8-102 Insert 2 1
The available source must be Class 1E or another reliable source. The available source must l be capable of supplying sufficient AC electrical power such that the redundant components are ' capable of performing their specified safety function (s) (implicitly required by the definition of , OPERABILITY). Otherwise the supported components must oe declared inoperable and the i appropriate conditions of the LCOs for the redundant components must be entered. l 6
P
' CHANGE ,
NLDBER ' JUSTIFICATION
' does not affect battery operation would not fail thelsurveillance
, acceptance criteria. This change is consistent with industry TraveleN TSTF 38. ,
3.8 42 An Action is added to the LCO for an inoperable automatic load
~
sequencer to immediately declare the affected DG and offsite circuit _ , inoperable. This Action is consistent with the CTS requirement that ;
~
when a Load Shedder and Emergency Load Sequencer is inoperable it is i
. appropriate to enter the Conditions for an inoperable DG and offsite i . circuit. .
I 3.8 43 For one DG inoperable while in MODE 1. 2. and 3. the requirement of the ; CTS to confirm the OPERABILITY of the [ turbine driven auxiliary ;
- feedwater pump is required to meet the redundant features requirement]
~
as part of ITS LC0 3.9.1. ACTION B.2 and a Note has been added to make [thisrequirementcie!.2. A ibenhi Nok he b% o.&k 4o L hea rch Ach n.2. mh c.1. H 3.8 44 - Not Gsed.
~
-.- . . -- .. - ._.. % 3.%. (-o l CW.~ --
)
~~T 3.8 45 The specifications for the power sources and power distribution systems
~
_ are revised to retain the CTS requirement.ftha one ain subsy tem) - L( shall OP E shut down. LCO 3.8. '3.8 and 3.8.1 are f r revi ire ~ rain
~-
7 7 ._~ to r or T ain to *one train of j,.; ,
, - T. " 2 equi ir to be E E. sev ral ati s. re erence to
! *' ne or re" dele ed. be nsis ent th t over 1 CTS
- i ~
equi t to trai be . L. T CTS 1r to ve cone ain o power sour sa power distr bution E pro des the i rt t oper e in saf na and miti te t seg es of st ated vents rinq[ shut . 7-j -
.... .._ $3".8-46[
~
Not applicable to Callaway Plant. (SeeTable Conversion 4 C
-I ((;.~g /g (Enclosure 68).
CRemove Lib thru Mk
- 3.8d7 Not applicable to Callaway Plant. See Conversion Comparisori Table (Enclosure 68).
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+, pass,-9, p.+ rs
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m.m m
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JUSTIFICATION FOR DIFFERENCES - TS 5 5/15/97
.- wa,.. , .. ..ee-air
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t ,
j , . o 1 , CONVERSION COMPARISON TABLE FOR DIFFERENCES FRON NUREG-1431. SECTION 3.8 Page 6 of 6 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.8-41 The phrase, "that could degrade battery performance " would Yes Yes Yes Yes be added to clarify the purpose of the battery inspection. This change is consistent with industry Traveler TSTF-38. . 3.8-42 An Action is added to the LCO for an inoperable automatic No No Yes Yes load sequencer to imediatffiseclare the aff4ted and - offsitecircuitinopera)(e. (ACnods Als1dnb C g 3.8-43 A Note is added to I ' 3.8.1. k d h M ate that [the Yes kes Yes Yes turbine-driven auxi iary feedwater is reauiral tn meet the redundant feat res requirement] ade ffG dndr/rM (Q 3 8 I-DICW) in Mode 1. 2. and consistent wit % Q gr5 geymg u.Q
- 3.8-44 Not used. N/A N/A N/A N/A
> 3.8 45 The specifications for the power sources and power No. See CNs 3.8 13 Yes Yes Yes
! distribution systems are revised to retain the CTS and 3.8-15.
j requirewntgh Qhe ys M @ . . at ( st sh[ p gpengov, (( q,h g Q3 9. t 0-ot cW w r ee TS i j . 3.8[7 The ITS LCO 3.8.3, CONDITION 8 requirement for a 'per- Yes No No No diesel generator" lube oil storage system 11 revised to I reflect the current design of a shared system between i i i units.
=
\ _
fCo oke (c) 04- ITS Tdla 3.S I**l . Osau d "fe a g, m.t. w a < % , co.,. a ts e 6 lut % thopy cou.J (s a.%g;,_u eN6k : 4 sp=EbL $ rwRy metemmtT( >
/ s if] '
- 3. 9 - S o ! Iriser' D ) 3 8N~I klC.
m 3* 3.:7 " CONVERSION COMPARISON T 'ABLE - NUREG 1431 ' , 3 5/15/97
, _ _ _ _ _ .. _ __.- __ _ _ _ ._ ____ _ _ _ _ . _ _ _ .. _ ._ m _ _
i i J l ADDITIONAL INFORMATION COVER SHEET ]
' ADDITIONAL INFORMATION NO: Q 3.9-1a APPLICABILITY: CP, DC, WC, CA REQUEST: CTS 3.9.1 _!
DOC 1-01-A - 1 ITS 3.9.1, LCO 3.9.1 JFD 3.9-15 a.' (Comanche Peak, Callaway and Wolf Creek) The CTS and ITS are proposed to be revised by adding "when connected" 1 preceding " Reactor Coolant System." The DOC provides a generic explanation, but it does not provide any specific technical justification for this addition. This revision is considered an administrative enhancement and a generic change to the ITS. Therefore, it must be reviewed and approved via the TSTF process l before it may be adopted as the standard ITS language. Furthermore, Diablo Canyon does not include the proposed addition, "when connected, " in its CTS markup. Comment: Either remove this item from the submittal and adopt the ITS language, or submit a TSTF for this generic change. Also, provide explanation why Diablo Canyon is not adopting the proposed language, "when connected." FLOG RESPONSE: (original)
-The proposed changes to CTS 3.9.1 and ITS 3.9.1 were based on traveler WOG-103, Rev.1. WOG-103, Rev.1 has recently been designated TSTF-272 and transmitted to the NRC in May 1998. The proposed wording in TSTF-272 was modified from WOG-103, Rev.1, and these modifications have been incorporated into the ITS and ITS Bases.
During preparation of the conversion license amendment request, WOG-103, Rev.1, was inadvertently omitted by Diablo Canyon. Diablo Canyon will incorporate TSTF-272 into the ITS and ITS Bases. i l FLOG RESPONSE: (revised) Based on the present status of the generic change process for the STS, it appears that traveler TSTF-272 will not be approved by the NRC in time to support the initial license smendments for the FLOG plants. In order to facilitate the issuance of these initial license amendments, an attemate approach has been developed which relies on the CTS, plant-specific information, and/or the NUREG but does not rely upon the traveler. This attemate approach is hereby provided as an interim submittal to allow issuance of the initial license amendments. The changes which rely upon the traveler can be i processed in subsequent license amendments following approval of the traveler by the l NRC. The FLOG believes that TSTF-272 would correct a potential literal compliance concem l while providing operational flexibility.. At this time, ITS 3.9.1 and associated Bases have l been revised to delete the incorporation of this traveler. LCO 3.9.1 and associated l Bases are revised consistent with the CTS and CTS Bases to clarify the description of l the refueling operation since the LCO is applicable only to the filled portions of the RCS ; i
and refueling canal that are hydraulically coupled to the reactor core. The boron concentration is established to ensure that the K, of the reactor core remains s 0.95 during refueling operations. Tt.is infers that the boron concentration limit is applicable j only for the reactor core or those areas indirect communication with the reactor core. ATTACHED PAGES: Att.15 CTS 3/4.9 -ITS 3.9 Encl. 2 9-1 Encl. 3A 1 Encl. 3B 1 Encl. 5A Traveler Status page,3.9-1 Encl. 5B B 3.9-1, B 3.9-2, B 3.9-3, B 3.9-4, B 3.9-5 Encl.6A 3 Encl. 6B 2 I I i i J 1 i
m 5, -\ d
*G
. RElyStay ,
4 ~
== . .. . ..... .
.,, l 3/4.9 REFUELING OPERATIONS 1.
t. 3/a.9.1 BORON-CONCENTRATTO_ . FOR OPERATION h 3sk-!a
'bMITINGCONDITI con concentration f all filled portions of the Reactor Coolant 1 - l-3'.9.1 The -
System and he refueling eene#shall be - d ati' et-t'"dty ed cae~renditf iad er*'ic'et
-- 1: 'a-
- r
=:; n th:t t' ::r: n:trkth: :f th: f:! b fe;
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- 8. ._ ._ . , . _ .
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- = :=natntf
= :f ;n:::r th= r :;u? t: 2000 ;; .
1-05-A APPt.ICAaItITY:.. 900E. _ . - 4% , h WD I**** ACTION: ~\ 4 a . With the requfrements. of he above specification not satisfied, immediately suspent all operations involving CORE ALTERATI0H5' or positive reactivity changes. and: initiate -d cent ce te-hf e . it ;--ster th= = ;;d t; ~~ ;;;. g .g(,.LS
;f : ::!ut:n :=u' '9; ;- it:- tri : :;ri! te ?OOO - 5 n: :r ' : ;;ui v--
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- b. U5.7 ' 'J.Es'.idic ,, bes ed suwed - +, 5,mk 4ks. % al aul i,, posi.4;
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SURVETELANCE EOUIREMENT
.. c_
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- a. 9.1. 2 The toren c:ncentration of the Reacur at Coolant System and the refueling 1-10-LG least once per 72 hours.
ta^r
-+ana 7' sna11 te cetarminec
- r. 1 T ".(p:te'n' t e w ni LQ a.9.1.3 Valves SG-V178 anc SG-v601 snal) te verified lected closec and secured in position at least once per 31 cays.
~
g d=**5 "h: n.:n = :N: ' i: - 'eufaed '- TOE ~ "- re * ' ': '- t' ::ct:- J.9 %
.;;;;1 _'in inc :::::' in :! :urt tr' t! 'n: tN= 'ti'yi:-'" : r in: x:: .c.:.c:. ,ake .5.uct. ;s .k.L .,.,
g _g,3
+ L 7a .s*.- q a 41,.nl f., c4 ck u h.ad<J
- v. Ivc..
+ s r. 9.9.t.2. iw4 k. c.. spl e!cd ad a. a c., .. h b is ea 4c re). 1 - CB-M CALLAWAY - UNIT 1 3/4 9-1
. - - . 2.
Description of Changes to current TS Section 3/4.9 F (Q %d;f This enclosure contains a brief description / justification for each marked up change to current Technical Specifications. The changes are identified by change numbers contained in Enclosure 2 (Mark up of current Technical Specifications). In addition the referenced No Significant Hazards Considerations (NSHCs) are contained in --. Enclosure 4. Only technical changes are discussed: administrative changes (i.e.. format, presentation, and editorial changes) made to conform to the improved Technical Specifications are not discussed. For Enclosures 3A. 3B. 4. 6A. and 6B. text in brackets "[ ]" indicates the information is plant specif1c and is not common to all the Joint Licensing Subcomittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location.
~
ME NUMBER NSEC DESCRIPTION - foDg, h3*q~}'
~
l-01 A Adds the filled portion of the refueling [ to the locations in which the boron concentration must"be ' maintained. -s o i ., . Q 3.3-l, These changes are consistent 5 tit s , current practice and current TS Bases. Therefore, they have no technicaT impact and are administrative. , 1-02 LG Specifies that the required limit 1 for the boron f- concentration will be in the Cor< Operating Limits Report. ' ;J. in accordance with NUREG 1431. T.ev.1. As the limits are currently in the COLR this change has no effect. In addition, the provision to maintain a uniform-concentration is discussed in the ITS' Bases.-This change removes details that are not required to be in the T5 to protect the' health and safety of the public while retaining the basic Timiting conditions for operation. 1 03 LG Not applica'ble to Callaway. See Conversion Comparison Table (Enclosure 38). 1 04 LG The specified limit on k,y s 0.95 is moved to the BASES: however, the limit is effectively maintained by the requirement te keep boron cencentration within limits which remains in the LCO. As noted in 102 LG above, the boron concentration limit will be mtintained in the COLR. This change removes details that are not required to be in the TS to protect the health and safety of the public while retaining the basic limiting canditions for operation. 1 05 A The footnote defining the " REFUELING" condition is not
, necessary because it duplicates the definition of Mode 6 in ITS Table 1.11. This change does not result in a DESCRIPTION OF CHANGES T'O ' CURRENT TS 1 5/15/97
k . </ - .
'l 2 ( 'f CONVERSION COMPARISON TABLE - CURRENT TS 3/4.9 Page 1 of 8 TECil SPEC CllANGE APPLICABILITY .
DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY IRJHBER DESCRIPTION n_ , e-(W Q'5.8-ld k,N Yes Atkis the filled portion of the refueling [ to the Yes Yes CTS wor 8p f4 l 01 A locations in which the boron concentration must be ~ do - 014w
- g ,7" Q 3.9-la.,l ggg,7pg Yes Yes Yes Yes l 02 Specirles that the required limits for the boron .
concentration will te in the Core Operating Limits Report. . LG ; in addition, the provision to maintain a uniform - concentration is discussed in the 115 Bases. l .i , Ho - Haintaining Instead of providing the tag rumbers of the valves used to l~No - CTS based on Yes , KNo-MaddaMJ ' ( k. 1-03 LG isolate unborated water sources, the function of the valves. licensed dilution \ ca tat C3 r1tMt4{/ valve nusbers in CTS l is used. The valve tag rumbers are moved to the Bases. b accident. l . 1 Yes Yes Yes Yes 1 04 The specified Ilmit on k,,, s 0.95 is moved to the BASES. ! tG i . No - not in CTS Yes Yes l 1 05 the footnote defining the *REftEllNG" condition is not Yes ' A necessary because it dupilcates the definition of Mode 6. Yes Yes Yes The requirements to initiate boration at a specified flow Yes . 1 06 t LS-1 rate having a specified boron concentration is replaced by " the more general requirement to initiate boration to ' restore the required boron concentration. Additionally. the action statement is revised to clarify that action is . applicable only to boron concentration No - CTS based on No - already part Yes Yes 1 01 A new ACil0H statement is incorporated that specifies the appropriate activilles if the isolation valves for licensed dilution of current IS , H , , unborated water sources are not secured in the closed i accident , position. I No CTS based on Yes Yes Yes 1 08 Separate entry into the action is a110wed for each
- 11 unborated water source isolation valve. licensed dilution' accident Yes Yes Yes Yes 1 09 the SR to verify reactivity coralltions is deleted. !
LS 2 5/15/97 CONVfRSI0t'l CollPARIS0!! TABLE - CURRfNT TS
fndustry Travelers Applicable to Section 3.9 S k3.N TRAVELER & STATUS DIFFERENCE 5 COHHEWS TSTF 20 Incorporated 3.9 2 NRC approved, TSTF 21 @ Incorporated Hot Applicable
-"CCcppnjg.de^ylkQLh Tg y,co, idir 23. Rev. -t- Incorporated 3.9 13 Traveler bracketed 3
ITS 3.9.2 and revised the Bases for 3.9.3. Bracketed Bases information from the traveler that is not 7p_y,93 applicable to a specific plant was not incorporated. M/C TSTF 51 Not Incorporated Hot Applicable wwe.l. Requires plant specific reanalysis-to establish j ggj decay time' dependence for fuel handling accident. TSTF 68. Rev'. 1 Not Incorporated Similar changes (Difference
#3.91) were incorporated into the ITS based on current licensing basis.
TSTF 92. Rev. 1 Not Incorporated Hot Applicable
') , ,
Not NRC approved as of traveler cut off date.
. TSir96,4v. / Incorporated 3.94 NRC approved.
1str 136 Incorporated Not Applicable l N g C aff n ve d -7p_lf, y@2 iSTF153 Not Incorporated Not Applicable .-t C ?pp-->d n .f *
. ..... ..- . . m... g 72-? 1-co.2
((dhd Incorpora fM f [Q8,9-7 / [ $[ [ncorporated) $4 l [
%J' Q 38~ k nh age ,7.nved do'h ,al cL y ,+ 2.! 5~un.,*cer. 7-wJ e,-ear '7. + and 2.5~
weo-e i-svo' s ed 5 &!
.r.9, n d.3 8
MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97
r
~
i i q g.9M _ Eoccn Concentration g@ 3.9.1 [(;\
^
3.9 REFUELING OPERATIONS . ja '34- lad 3.9.1 Scron Concentration yi / , Q 3 A-la l t LCO 3.9.1 Baron concentrations o th Reactor Coci ant System. "..~ . . '.. . '. ..,- f I fyy gj i . eener.- maintained andp'within the limit specified in the COLR=--2rcT
* ,'the _# ref 3
tM- 6. Awn.& a.ue.ss b i ! l
~
APPLICASILIT/: MODE 6. - ** "5N 3 a - -- w _ ._
..,. m nu w 2 ww= g_ A E 9 -la. l RETIEirME"#* JMi's:;nortme.tMentnyant'@B00mmmEGesch ;a,gn I pe:;mCE N a u -+ r: -_ . __ _ ;_ -
~
C6SDITION REQUIRED ACTION CCHPLETION TIME A. Scron concentration A.1 Suspend CORE ALTERATIONS. Imediately not within limit. A.2 Suspe'nd positive-- -
-Imediately reactivity additiens.
h.EQ A.3 Initiate action to l Immediately restore bcron ' concent. ation to within limit. SURVEILUNCE REQUIREME.YiS SURVEILLaCE rREQUENCY SR 3.9.1.1 Verify boren concentratien is within tne limit 72 hcurs
- spe.1fied in the COLR. 50 MARX UP OF WOG STS REV 1 UE' REG 1431) 3.9 1 5/15/97
i h V Boron Concentration J
\f B 3.9.1
- . . ~ h htP y Av 1 .
d
.u
.! B 3.9 REFUELING OPERATIONS Q 3.Ma, B 3.9.1 Boron Concentration . o t1 nt BASES I -
i . BACKGROUND, 1 The limit on the boron concentration 4 ofFthe Reactor Coolant ! System (RCS). th; pfalia; sa;1. ad th; refaling svity g z.inauenna2 Door-during refueling ensures that the reactor hg#[y ' gM t j remains _ subcritical during i40DE 6. Refueling boron concentration is the soluble boron concentration in the coolant in each of - these volumes having direct access to the reactor core during refueling. j i g N4p h The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of ! s the coolant in each of the volumes. The refueling boron j g,\u concentration limit is specified in the COLR. W WrG j 'Mi+Tr ualahV ' i- dii1J PMidgMildlini_itt 4 6MeV is%h]A*w 8 3 l ShE%h 1
- 3Ect'tpDERIR5ffmERTstW Pter*
l 5.;ada; ;; z;a;^; :.r; ;, zl'.l: L^,z ;;.;; ;:.^;:l:,; l,7 ;,-::.^ ::
^ ^ ^
nht:i = :=r:ll Cr: rn?.i"it: :' 6 ; 0.00 d;rir,; ';;l hadlias with satr;1 red; c.r.d fal us hlis a;;. ;d t; b; in th; e t -dars s,.'igur;tien 'is;t a;;; tis s;tivity;
;11;ad by plat piesdera.
GDC 26 of 10 CFR 50. Appendix A. requires that two independent reactivity control systems of different design principles be provided (Ref.1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. The i Chemical and Volume Control System (CVCS) is the 15151 system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration. The reactor is brought to shutdown conditions before beginning ;
. operations to open the reactor vessel for refueling. After the ,
RCS is cooled and depressurized and the vessel head is unbol"ed. ! the head is slowly removed to form the refueling cavity. W ! FEfDE13h1Tav@hMFf1 tid 5Lrdt't2rJorm:T, nesrdfiiETTR3555_1,,,I' The. r;f ;11a; snel .r.d th; r;'slin; svit., er; IBiica'lDM l 1 (continued) i MARK UP OF NUREG 1431 BASES B 3.9 1 5/15/97 i
Boron Concentration B 3.9.1' O y '.
.a BASES-BACKGROUND iflilEUlErijRi51Hf then flooded with borated water from the (continued) refueling water storage tank through the open reactor vessel by grnity fadi .; er by the use of the Residual Heat Removal (RHR)
System pumps _52p'IR1G*5'H5ftlR
. The pumping action of the RHR System in the RCS and the natural circulation due to therman on ting heads in the reactor vessel and refueling eevdy M mir the added concentrated boric acid with the water in the refe.iing eenet- M The RHR System is in operation during refueling (see LCO 3.9.5. " Residual Heat 4 Removal (RHR) and Coolant Circulation High Water Level." and LCO 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation -
Low Water Level") to provide forced circulation in the RCS and assist in maintaining the ORWAYS boron concentrations in the RCS. tre refalia; ;;.ac.l. and the refueling eevey .
.v above the COLR limit. Ymngn m wgmE a.-< ..h.e S ..s c o
'i Ei10TMse21J.aMIMG i, EQ e- @ iT M 4 m + v f +1
....r. w. e g .- % r y m. g.h g .y & .n 1.je g 2.p 1, i9tEPi4M%%D(Sy&M==NmrPDWePr 9 APPLICABLE :: aria; r;falia; ep;r;. tier.;. the r;;;tivit3 cc,r.ditier ef tra Gr; SAFETY ANALYSES i; cen;i; tat .;itt tra initi;l _..dities nand fer the berc,a dilutica ecident in tra exid;nt ; sly;i; ; .d i; sa;;rs; tin fer "00 C. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.
The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including full core mapping) ensure that the k,n of the core will remain s 0.95 during the refueling operation. Hence, at least a 5t ak/k margin of safety is established during refueling. ERR 5F A5MTtBgit'fvWiltDRtt515NiBUNR tilEENRYt5c575 rte 59RERHfl5ErlHMfDEMIEDU51gTJETE ttt!N
., assumpe.'UE
.During refueling, the water volume in tre ;p;at f ;l pc~1. tra tr;..;f;r sal. tfiEil5Ef0EMINRl5T tra c;falia; sal, tra refali .; ;;vity. and the reactor vessel form a single mass. As a result, the soluble boron concentration is relative 1 the same in each of these volume Og 4 3,q L
&in Nimb" td< vasd. (continued)
MARK UP OF NUREG 1431' BASES B 3.9 2 5/15/97
u. Baron Concentration
- B 3.9.1 4
.. . g
, p,.:a.d... ~r. BASES } ' APPLICABLE -The-limiting- boron dilution accident analyzed occurs in H0DE 5
~
i r SAFETY ANALYSES'~ J
' (Ref. 7.)~~h-dt.tcilcd discussica c' thic cvcat is gcvided in (Cont 2l nut'U) ' '."."." '.'_~ e.,i. ,,
.,c, e' O. . . e 9- e m o v- =
i
. -_- .. a c.,u.t.r.mr.u.. t u.m. .r..n. .t sti'Hp b W r.tSisjf: 4et y *.t . f C - M* ga(vs p
e,r.,.y us 0;. s ww 4 . Q
. ia. '^.J' T, cf2... b a[e,;.je g ii. g e'p l 'sH r o v ' '(sC y ?.);: p .. i'4L L R y J. .:'. .-
_ q q e.,F .u.:.;. W.g;e C h i.- 4.:. 9. n m \M.?'} % '^ WJ r.n t%;s Am_ ' - . s i - n . . 's . ad:7, q s ~ w ey . Qi ..,etsy q sp. gs. .io. . r.gfah .J; uwm: . .w ...C, .sy .. .c;- :p p . . , ' . g ;,4
~
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g
. 'Q ;;:
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j .g* ._.;__.__. . , , , .,.. , , . , . _ a . : igg b fth ;N Y= E- y _l
,. 6 .e .'
( a sf . i . si . hG t ' '. _ '* f
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,5 1..M 1 .a c.3 e c 3;. g...'g g. q g- .
q c-y .g .w y,7 y.y ; a
. k"' ' J&'C'a.' 3 b_Y < ! " 'b '
.'.( '.I.he iE ~ 4 ' PS'S ' % Mhk '#N*M,
! -p. v. g . . . q .. n . . . ; u .
- r. v , .4 c. . , , y c.n s' -~
+g 1[
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,c g g nu Ar4+=rf.jg a ,
1 1..,c _ 9,..f.y. W .y - i_ I j g 3.h-/c The RCS baron concentration satisfies Criterion 2 of thc CC li Pciicy Stct;;.c.-t. 4- M:- M.y 6, m u - l - 1
'ECO The LCO requires that a minimum boron oncentration be maintained tt in 4rw&yhu.le 9.v r the RCwwn- _ ,-~ a s_--
-- crc l
._d. p~
r m"-
.--__~ . -- - . _. __ . . . . , _ . . y q3,9.(
m __,._,,__ _ . . w * . 6 y) 4 6 400C 0. The boron concentration a
'f (antinued)
MARK UP OF NUREG 1431 BASES B 3.9 3 5/15/97
--. - . - . -. .-. _. - _ - . . ~ - . .. . . - . - - . - . . - - - . - - . -
. . . . . . . - ~ .
Baron Concentration B 3.9.1 r.c e . Eh .> . ' BASES
-w. .!
LCO - . . ~~~Timit'~specified in the COLR ensures that a core 4, of s 0.95 is (continued),, maintained during fuel handling operations. Violation of the LCO 1
- ~.- .. - could lead to an inadvertent criticality during MODE 6. :
l APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron i concentration ensures a (r s.0.95. Above H0DE 6. LCO 3.1.1,
" SHUTDOWN MARGIN (SDH) --T,,, : 200 *I. '" and '_C0 :.'. 2. ~5"'JOO'e l
.".^ RC!" (S0") T.,j e.. .,
,, g.3 . : 3 s . % ,.._ g.a.p .,m m w a f. z .. ensure
~ that an. adequate amount of negative reativity is available to !
-- shut down the reactor and maintain it subcritical. !
, C G 2 _~_ Q L 9 ._def;G.. - .
.; , .. r. .;; . . - . _; ., . ,.r_ e u . g vr.c . .w A
.,u ~e . '; sep . . _ . , . :Q ' St..;
- y. ' . % . ~.5 . u .
.,"'**~-*- '\ ,, c '.. ..; ;,: a ' s .g .o - 7 :.; i . 1' . %- 1 . - so'ef @ 4. ; . :.a= 'i q
{.'.:
-- ,- i ;_,, 17f;. .. p Q . . a~ i. - ....:i.... .a 4 g (a,4.y . '{ea..-
w wy a 3.cs a.n p .e .- ..r. ~.a,%. a x.m .pis ;.~ S,. . . . , _ we .xe y w t. .g.
.=-
. =- O -
.2:. ..;;-
%' ACTIONS =- "- K.I"and A.Z
~
~- . . . . . . -
~ ~
Continuatio'Tn of CORE ALTERATIONS or positive reactivity additions
' ~(incTuding actions to reduce boron concentration) is contingent I e I upon'. maintaining thunit in comoliance with the LCO. If the 'ff h' l# ..._
._fxicorEconcentration of any coolant volume in%he RCSam 6less. th;than .xfucling its limit, canel, er the refueling ewer @ IONS or
.i s p,g,(
f0(Mg all operations involving CORE ALTERAT positive reactivity additions must be suspended imediately.
~3, Suspension of CORE ALTERATIONS and pcsitive reactivity additions shall not preclude moving a ccmponent to a safe position.
Y h U l AU h eW In' addition to imediately suspdhding CORE ALTERATIONS or p hg positive reactivity additions, boration to restore the concentration ^ must be initiated imediately. g)d (continued) MARK UP OF NUREG 1431 BASES B 3.9 4 5/15/97
Doron ConcGntration B 3.9.1 I BASES ACTIONS A l (continued) In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.
~
SURVEILLANCE REQtJIREMENTS SR 3.9.1.1 fnf' # 1 This SR ensures tha* the coolant baron concentration in M6 %yp I TUlsdVortidrisy6f he_RCO? the rcfucling cenei and the 13 i rcruc$ricivitz* -_ie".^~ M m;...a r"cr --2:x;O EF* 5tf"4=W- di-cD.W.# d;..i-57i-U2iB n- [ f @F
) ~
is within the COLR limits. ~The boron concentration of the coolant in each teguir~ ell volume is determined periodically by Nh;i S,p ! chemical analysis. A minimum Frequency of once every 72 hours is a reasonable amount b9k1 U J q.l of time to verify the boron concentration of representative !.' C.h samples. The Frequency is based on operating experience, which t= has shown 72 hours to be adequate. { REFERENCES 1. 10 CFR 50. Appendix A. GDC 26.
.o
- 2. FSAR. Chaptcr [15]. Chafitiirgnme~R16ny5?4! .L t
Mrjded_nt',yfito; tac 1 fitTDpei.ItJjjRJ.TcWii' s i3N6T3HPE:30; E Ci))awayWit3 Hat _ed tHaYcM312".i1995; F. C811awa.yZPgnQi@ue~st*f6r3esWtM]ZUZ[
)
MARK UP OF NUREG 1431 BASES B 3.9 5 5/15/97
l \ l l l CHANGE NUMBER DESCRIPTION l fphy'; if the baron concentration limit for MODE 6 is not met. (See the LS 1 NSHC in the CTS Sect 4.0, ITS Section 3.0 package.) I __Q 5 9 .t4 - ..ulat. (oA-f > -7 ! ! , 3.9 15 L .9. h b n vis in ccorda ce wi h tr ele V' " 'us t ! ! ai th on nce rat ~on lim t ply to th ef lin l 1] r f1 de rea when ese reas e t co ect to he R h'a e is a ep e be use e on ncen ati t i i i ed o e re hat he tor ai sub ri ' 1i MODE : e, na as ntai ing n uti are isol f om t g'3,hlu R , t' 1 f b n dilu ion 1 . Th refor , t .i no o ace li t on once rat on ~ hese are whe ey ae t onne te t RCS. I han e is e ent ith t e i nt th Spe ifi ion, des 1 in t e Base , nd eld inates
, re tri tion ha have no f a safety.
u . ,-
~
n rm v .~u 7 H S G 9 y- x 3,9- l6 Zn.w.}- 6A- b ) '
.Q sa-7 :
1 E.' * - l l 1 i 1
-4 l
i l n.Jarr * ,. JUSTIFICATION FOR DIFFERENCES - TS 3 5/15/97 l
I _: O 3.9-1a i ! INSERT X i' 3.9-15 LCO 3.9.1 is revised consistent with tiie CTS and Bases to clarify the description ; of the refueling operation since the LCO is applicable only to the filled portions of l
- the RCS (and refueling pool] that have direct access to the reactor vessel. The boron concentration is established to ensure that the K., of the reactor core remains s 0.95 during refueling operations. This infers that the boron concentration limit is applicable only for the reactor core or those areas in direct communication with the reactor core.
s l I l 1
'i I
( -.
\, '
.[;.7) t::;. -
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG 1431, SECTION 3.9 Page 2 of 2 DIFFERENCE FROM NUREG 1431 APPLICABILITY NUMBER DESCRIPTION DIABLO CN1 YON COMANCilE PEAK WOLF CREEK CALLAWAY 3.9-8 The NOTE of ITS 3.9.5 is expanded to incorporate current Yes No not in CTS No - not in CTS Ho - not-in CTS licensing basis allowing the RIR pump to be removed from service for less than or equal to 2 hours per 8 hours for leak testing of the RIR suction isolation valves. 3.9-9 The survelliances of ITS 3.9.5 and 3.9.6 are modified to Yes No not in CTS No - not in CTS No - not in CTS incorporate current Ilcensing basis of two RlR flows dependant upon the number of hours the reactor has been subtritical. 3.9 10 ITS 3.9.7 Appitcability is being revised to delete "During Yes Yes Ye a Yes r e . CORE ALTERATIONS. except durt 1 tc 1 , unlatching of t 3 ntrol rod driv _e hafts.' i b iy bI is e ir_ it to ra or ve e 3.9-11 c -d w h ra- 1 IF 6. he CD s if ed S Mf_ h y M mi ta on pa th pr i re a Afp y A/ A t i n t t t de mo et be ns _r . ls/ist tly con o) . f & USEb O 3'0" 7 A note is added to the Appilcability of LCO 3.9.6 fes Yes Yes Yes 3.9 12
~
indicating that entry into a HDDE or other specified condition in the Appilcability is not permitted while the LCO is not met. Yes No No No 3.9-13 In accordance with DCPP current TS. LCO 3.9.2 would not be used. A note is added to the Applicability of LCO 3.9.1 - Yes Yes Yes Yes 3.9 14 indicating that entry into M00E 6 from MODE 5 is not permitted while the LCO is not met. t1 1 It no a aPa'@ gg g bn ca o AIMERT ts-3h -> 3.bl6 L'f*f 68-h G3T-M CONVERSION C0tlPARISON TABLE - HUREG-1431 5/15/97
Q 3.9-1a INSERT 6B-3b TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO COMMANCHE WOLF CREEK CALLAWAY. CANYON PEAK .; 3.9-15 LCO 3.9.1 is revised consistent with the CTS and Yes. -Yes- Yes Yes Bases to clarify the description of the refueling operation since the LCO is applicable only to the filled portions of the RCS [and refueling pool] that ! have direct access to the reactor vessel. i i t i
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 5.5-2 APPLICABILITY: DC,CP,CA REQUEST: Difference 5.5-14
- Comment: WOG-85 has not yet become a TSTF. Use current ITS.
FLOG RESPONSE (original): WOG-85 has been approved by the TSTF and is designated as TSTF-237. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-237 was modified from WOG-85 and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes proposed by this traveler. For Wolf Creek, this change was approved by the NRC in Amendment No.106, dated June 24,1997. Therefore, the wording in ITS 5.5.7 is consistent with Amendment 106. FLOG RESPONSE (supplement): Based on a telecon with the NRC staff on January 7, 1999, the NRC staff approved TSTF-237 for the FLOG. The approval for the FLOG was
. contingent upon a revision to the DOC that addressed the RCP tlywheel material.
Enclosure 3A, will be revised to include a description of the RCP flywheel material, and a statement that the items specified for accepting the change in the conclusions section of the NRC staff SER on WCAP-14535 are met. ATTACHED PAGES: Encl. 3A Page 6
CHANGE IMBIS 15EG DESCR1PTU)B %Th Flywheel Integrity." The proposed exception to the recommendations of Regulatory Position C 4.b would allow for an acceptable inspection method of either an . ultrasonic volumei.ric, or surface examination. The acceptable inspection method would be conducted at ten
-~
year intervals coinciding with the Inservice Inspection schedule required by ASME Section XI. This change is consistent with the NRC Safety Evaluation Report associated with WCAP 14535. Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination.
.Irisert 3A-Q 02 18 A Revises the Radioactive Effluent Controls ogram dose rate limits released to areas beyond the ite boundary to 4 5.7.-l }
reflect new 10 CFR Part 20 requirements.ltonsisten with
~
FNRC lettgr dated 7/ZB/95/(Christopher 1. Grimes Owners t Groups). Addifionally./the NRC [ssued a draft seri & Lette ' i199 which p upused C anwe= to-t dard Tech ical S cificati ns. Thi change is cone ~ stent with l the draft neric Le ter and REG 1431. Rev. I as amende by a propu>ed Lraveter te ref ect chang Mstent wi I (1 CFRParb20. I 02 19 LS 2 Consistent with NUREG 1431. the surveillance interval for verifying that other properties are with limits for ASTM 2D fuel oil is changed from "within 30 days" to "within 31 days" after obtaining a sample. The fuel properties that can have an insnediate detrimental impact on diesel combustion. (i.e. , API gravity, kinematic viscosity, flash point and appearance) are verified prior to addition to the storage tank. The "other properties" may be analyzed after addition to the tank. The 31 day verification interval for these properties is acceptable because the fuel properties of interest even if they are not within their stated limits, would not have an immediate affect on diesel generator operation. The CTS 30 day verification interval was probably chosen because it was a convenient time interval for sending the sample and receiving the results from the laboratory selected for testing. NUREG 1431 has selected a 31 day testing interval . The 1 day increase in the interval would not l _ ._ . _ . _ _ - have a significant affect on the acceptability of the j diesel fuel oil. 02 20 A Consistent with NUREG-1431. Rev. I and traveler TSTF 118. - add the statement that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program frequencies. These sentences provide consistency with the current application of these requirements as l provided in ISTS 5.5.6 and ISTS 5.5.11. Amendment [112] l moved the Diesel Fuel Oil Testing Program to Section
6.0 DESCRIPTION
OF CHANGES TO CURRENT TS 6 5/15/97
y , t INSERT FOR Q 5.5-2 Enclosure 3A, Page 5, insert 02-17-LS-1
- Integrity of the RCP flywheelis assured on the basis of the use of suitable material, adequate design and inspection._ The flywheel consists of two thick plates bolted together. The flywheel material is produced by a process that minimizes flaws in the material and improves its fracture
- toughness properties, such as vacuum degassing, vacuum melting, or electroslag remelting.- Each plate is fabricated from SA-533, Grade B, Class 1 steel. [Callaway Plant] has confirmed that the flywheels are made of SA-533 B material and that the items specified for accepting this change in the conclusions section of the staffs SER on WCAP-14535 are met.
l l- ! l . l ! i f 5
, 3 I
l l-l. l T _ _ . - - - i}}