ML20095C948
| ML20095C948 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/31/1992 |
| From: | Hollomon T, Joshua Wilson TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9204240220 | |
| Download: ML20095C948 (10) | |
Text
.
9 s.
(* 4 k
3._..s,. y w _ , i u .. . ., n w . % . i. ,,
JL mon
'v s. t:I4 t 4 i i I 4 ,f I-Aptil 15, 1992 U.S. Nuclear Regulatory Connission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority )50-32b SEQUOYAll NUCLEAR PLANT (SQN) - MARCil 1992 MONTi!LY OPERATING REPORT Enclosed is the March 1992 Monthly Operating Report as required by SQN Technical Specification 6.9.1.10.
If you have any questions concerning this matter, please call M. A. Cooper at (615) 843-8924.
Sincerely, -
J. , Wilson Enclosure cc: See page 2 qV 3 9204240220 920331 I gDR ADOCK 05000327 PDR
/'/}7 U d i 1
s
.- q
~
e: ,
-U.S. Nuclear Regulatory Commission ",
Page 2-April-15, 1992 f i
ec (Enclosure):'
INP0 Records Center l Institute of Nuclear Power Operations 1100 circle 75 Parkway, Suite 1500 E Atlanta, Georgia 30389 r
Mr. D. E. LaBarge, Project Manager ,
U.S. Nuclear Regulatory Commission '
One White Flint, North 11555 Rockville Pike l Rockville Maryland 20852
- Mr. Ted Marston, Director Electric Power Research Institute P.O.-Box 10412 ,
Palo Alto, California 94304 NRC Resident Inspector Sequoyah Nuclear Plant ,
2600 Igou Ferry Road
- Soddy-Daisy, Tennessee 37379 Regional. Administration
-U.S. Nuclear Regulatory Commission Office of-Inspection and Enforcement Region II ,
101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 2 f -Mr. B, A. Wilson, Project Chief L U.S.- Nuc1' ear Regulatory Commission >
Region II
-101 Marietta: Street,-NW,. Suite 2900 Atlanta, Georgia 30323 l -
Mr. F. Yost, Director Research Services sUtility Data Institute
-1700 K Street.-NW, Suite 400 Washington, D.C. 20006 h
TENNESSEE VALLEY AUTTIORITY NUCIJAR POWER CROUP SEQUOYAll NUCLEAR PIANT MONTillY OPERATING REPORT IV Tile NUCLEAR REGUIATORY ComISSION y itARCillD2 4
UNIT 1 DOCKET NUMBER 50-327 LICENSE NUMBER DPR-77 UNIT 2 DOCKFT NUMBER 50-328
- LICENSE NUMBER DPR-79
OPERATIONAL SUtMARY MARCil 1992 UNI L 1 Unit 1 generated 499,580 megawatthours (MWh) (gross) electrical power during March, with a capacity factor of 57.79 percent. On March 6 at 1622 Eastern standard time (EST), a power level decrease was initiated because of delta T/T avg probleeas associated with cable-induced spiking of the overpower and overtemperature setpcints. At 1822 EST on March 6, Unit I reactor power was at approxime*.ely 81 percent. Unit i reactor powcr remained at 81 percent until 103G EST on March 8, when a power level increase was initiated. Unit 1 was again operating at 100 percent reactor power level on March 8 at 1725 EST.
?
On March 18 at 2048 EST, Limited Condition for Operation 3.6.5.3 was entered after 11 of 48 ice condenser icwer inlet doors were found to require excessive _
force to open. On March 18 at 2210 EST, shutdown of Unit I was initiated.
Unit 1 entered Mode 3 at 0247 EST.
On March 19 at 1347 EST, with Unit 1 still in t; ode 3, it was determined that there was a feedwater Isak on the steam generator (SG) No. 3 line that could not be repaired at this temperature. Unit I was taken to Mode 5 at 0205 EST on March 20 to allow further investigation, testing, and maintenance. After Unit 1 was taken to cold shutdown, an inspection of the feedwater line identified a circumferential crack at the transition piece between the SG's norr.le and feedwater line. Radiographic examination also revealed transition ,
piece cracking on Unit 1 SG No. 4 and Unit 2 SG Nos. 1, 3, and 4. A decision was made to replace the transition pieces on all eight Units 1 and 2 SCs.
Unit I remained in Mode 5 at the end of March.
UNII_2 Unit 2 generated 271,110 MWh (gross) electrical power during March, with a capacity factor of 31.36 percent. Unit 2 was at approximately 79 percent -
reactor power level at the beginning of March and was in coastdown to Unit 2 Cycle 5 refueling outage. On March 13 at 1700 EST, a power level decrease was initiated to bring the unit of fline f or the out age. Uni' 2 entered Mode 2 at 2125 EST and Mode 3 at 2130 EST on March 13. Mode 4 was entered on March 14 nt 1101 EST, and Mode 5 was entered at 0902 EST on March 15.
Unit 2 entered Mode 6 on March 20 at 2158 EST, when the first reactor vessel head bolt was detensioned. On March 23 at 0550 EST, the reactor vessel head was lifted and inspected. Core offload began on March 26 at 0710 EST; core offload was completed on March 29 at 0410 EST. Unit 2 remained defueled in "no mode" at the end of March.
E0WER-DEERATED .RELIELVALVESlPORV1 AND_SAFEILNAP. ;S. St.M1ARY There were no challenges to PORVs or saf ety valves in iiarch, i
l
- AVERAGE DAILY UNIT PWER LEVEL DOCKET NO. __.10-327
_ UNIT No. __One DATE: _34-01-92 COMPLETED-BY . L ._,lx_llol l o m an TELEPH0!!E: Ibill_S43-1328 MONTil: MAR.Cll_19.92 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY OMe-Net) DAY _DNe-Neu 1 1142___._.._. I7 - - . . . _ _ . .11A1 _ __
2 _- 1141 _
18 1122 3 11A3 19 20 _
4 1105 20 _ -L 5 L12h _
21 _ _ . . _ _ . _ -9 ___
6 _ lQS6 22 _ __ -l .__ . . . _
7 939 23 -7 8 _.1D28 24 . -j 9 _llh2 25 -7 10 1143 26 _ -lh 11 11h3 27 -2 12 1144 28 -14 13 1144 29 ._
-5 __
14 1142 30 --i 15 1119. 31 __ ;-L__ _ .
16 1137.
l
,- o AVERAGE DAILY IINIT POWER IEVEL DOCKET NO. _50-328 UNIT No. _.Iva DATE: J4-0_3-22 COMPLETED BY: Ii l _}{allomon TELEPIIONE: .(6.1hLE43-Z528 MONTil MARCIL129.2 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DM (NWe-Net) DM . ._ONe.-Ne i1 -
1- __ _. 8 2 2_ _ _ . ..__ _ 17 .._.__.___._._-l__._...____.
2 885 18 -4 3 872 19 _. -lb _.
4 Sb8 20 __
--14 5 851 21 __ -l__
6 SM 22 _ _______n 3 ____
7 830 23 __
-9 8 8h1_ 24 ._ - R_ _
9 S34 25 - Z_ _ _ _ _ _ _ . _
r 10 __B29 26_ -1 11 821 _ 27 -12
-12 814 28 -16 _
13 bi6 .
29 - - - -9 --
14- -- L9_ 30 -7 ___
15 -lb -_ 31 _.
-9 ._
'16 -9 l
- . . . . . . . - - . . - ~ , - . - . -. - . - . - . . . _ . - .-- - -> - - - - - -
n.
OPERATING DATA REPORT DOCKET NO. ._50._327 DATE JphR22_
COMrtETED av Ladiojicopa__
TELEPHONE [6JJLQ43-1528_
OfLRAJJNC STATUS l Notes .l 1 Unit Name: _ legynygitjJnLLQnt_ _ . _
l l 2 .' Reporting Period:: tieritL1992___ l l
- 3. Licensed Thermal Power (HWt): _341hD ___
!. l 4 .' Nameplate Rating (Gross MWe): ._1Z2(L6 l -l
- 5. Desinn Electrical Reting (Net HWe): 11400 l l
- 6. Maximum Dependable Capacity (Grous MWe): _llf2 A l l
- 7. Maximum Dependable Capacity (Net HWe): _.112130 1 _j
- 8. If-Changes Occur in Capacity Ratings (Item Numbers 3 Through 7) Since Last Report, Give Reasons:
- 9. Power Level To Which Restricted, lf Any (Net MWe): N/A
- 10. Reasons for Restrictions, If Any: NB This Month Yr-to-Date Cumulative
- 11. Hours in Reporting Period 744 2.184 94249
- 12. Number of-Hours Reactor Was Critical 43h3_ 1.8&L 48 222d_
- 13. Reactor Reserve Shutdown Hours A 0 0
.14 _- Hcurs Generator On Line _ 434.2 1.B]A2_ 47 145.3
- 15. Unit Reserve' Shutdown Hours _ Q20 -0 Q__
- 16. Gross Thermal. Energy Generated (MH) 1.440.810.11_ 5,1ZL2LB_ Jh5JBL770
- 17. Gross Electrical. Energy Generated (MWH) _ d9L 580 _2.,D6L71H_ _32,1.3L264_
- 18. Net Electrical Energy Generated (MWH) 41LIZ2 1.9E60L _ 50.522 339
- 19. Unit Service factor _
58.4 85.8 SD.7
- 20. Unit Availability Factor _
58.4 _H5 B_ 5.G 11_ -
- 21. Unit Capacity Factor (Using MDC Net) 57.5 81.1 41,.g_
- 22. Unit Capacity factor (Using DER Net) ' 5fL2 _ 79.3 46.7
- 23. Unit Forced Outage Rate 41.6 _ & Z_ - 4.pd_
'24.. Shutdowns Scheduled Over tient 6 Months (Type, Date, and Duration of Each):
PS if Shut Down At End Of Report Period, Estimated Date of Startup: J pill _lL _1992
- * * '?i4@ t--'. 1- W N r- W' #
.- ._._m_.- . _ _ _ _ - _ . . . . _ _ _ _ _ _ . - _. _ . . . . ~ . _ _ . . . _ _ . . . _ _ _ _ ~ . _ _ _
h l- r 1
-OPERATING DATA REPORT DOCKET NO, _50.-32S_ _ ;
DATE .JedJ292 _ '
COMPLETED BY L W inl]Dmpft_ .
TELEPHONE M151_04A7528 QEIJRUt!G 51AM l Notes l 1, Unit Name: JtRug,yah Unit Two l l
- 2. Reporting Period: _fia.tsh 1992 l l
- 3. Licensed Thennal . Power (HWt): 3411.0 l l
.4. . Nameplate Rating (Gross MWe): 12202ft l l 5; Design Electrical Rating (Net HWe): 1140.0- l l
- 6. Maximum Dependable Capacity (Gross HWe): JM2J) l l 7 Maximum Dependable Capacity (Net NWe): .l]22 d l_ l
- 8. -If Changes Occur in Capacity Ratings (Item Numbers 3 Through 7) Since Last Report, Give Reasons:
- 9. Power level To Which Restricted If Any (Net HWe): _N/A
- 10. Reaso:.s For Restrictions If Any: N/A This Month- Yr-to-Date Cumulative
'11, Hours in Repnrting Period 744 2 dB4_..,. 86.209
- 12. Number of Hours Reactor Was Critical 309dL _ lllLi, 50,22L _
- 13. Reactor Reserve Shutdown Hours 0 0 0
- 14. Hours Generator On-Line' 30L1__ 1.70L5_ 43L2.iZ1 15, Unit Reserve. Shutdown Hours 0.0 0 0
- 16. Gross Thermal Energy Generated (HWH) . __293J)25.1
_5216L13241_ _15LDBS 045
- 17. Gross Electrical foergy Generated (HWH) 271.11G 1.76L256 12 & 0.557 18.-Net Electrical Energy Generated (HWH) _ 25.4._.05R 1.691.711 19, Unit Service ractor
_.5DE6A7">
41 5_._ 78.0 57.7
- 20. Unit Availability factor 41.6 78 JL 57.7
- 21. Unit Capacity Factor (Using HDC het) 30.5 _69JL SL4_
22; Unit; Capacity Factor (Using DER Net) _ _ 22,6_ ,_ _ GLj_ 51.2 23 Uni _t Forced _0utage Rate Q.0 2.6 35.3
- 24. Shutdowns Schedule 6 Over Next 6 Honths (Type,.Date, and_ Duration of_Each):
- Unit 2 Cycle 5_ refus1}ng_. gut. age Aegan. March 13. 1992.
- 25. If Shut'Down At End Of Report Period. Estimated Date of Startup: _liex_lL_.J232
a DOCKET NO- 50-31/- '
UNIT SHUTD0%45 AND POWER REDUCTIONS UNIT NAME: _ On e . .I DATE: 04/03/92 . .
. REPORT MONTH: March 1992 ' CDMPLETED BY:T. J. Holloman TELEPHONE:t3151-343-7523- .I r
Method of' Licensee Cause and Correctiva- i Duration Shutting Down . Event "3ystem Component Action to i No. Date l Reason2 Reactor 3 Report No. Codeb - Prevent Recurrence Type (Hours) Code # f 1
3 3/6/92 F N/A B 5 N/A N/A. N/A- On 3/6/92 at 1622 EST. Unit 1 [
power level was reduced to '
approximately 61 percent .because of delta T/T problems.' A .
power level'inErease was ' initiated !
.s at 1030 EST on 3/8/92. Unit I was at 100 percent on 3/8/92 et. :I 1725 EST. k I
i 14 3/19/92' F 309.8 .8 1 327/92007 BC DR On 3/19/92 at 3210 EST, Unit L1 f was taken offline after inspection of ice . condenser lower inlet doors showed 11 of 48 doors reovirei excessive force to open. Unit I entered Mode 3 on 3/19/92 at 0247 EST. :
l Dn 3/19/92 at 1347 EST, with 'i Unit 1 in Mode 3, a feedwater -
leak was identified on the 5G
- 1 I
No. 3 feedwater line. Unit I was taken to Mode 5 at 0205 EST on
- 3/20/92 to allow further investi-gation, testing, and maintenance. .
A decision was made to replace the '
transition pieces. j i
t 2 g,,3,n 3 #
I:F Forced Hethodi Exhibit G-Instructions j 5: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data j B-Maintenance or Test 2-Manual Scram Entry sheets for Licensee C-Refueling 3-Automatic Scram Event Report (LER) File
[
D-Regulatory Rest ruction ' 4-Continuation af Existing Outage (MUREG-1022) .!
E-Operator T aining and License Examination 5-Reduction '
F-Administrative 9-Other S
G-Operational Error (Explain) Erhibit I-Same'Scurce H-Other (Explain) g- . .,g. - _ g -7wi- -_m-_
, ;i)! !.;[> f(Ef5 !i t! IhE r tl A f f t! ith6i!i t'It8l -
! { t tf!II=' }; e i'
+
I jIo 4N 5 h e g t . s r IM%7
* - a w2 a i3 4
- i me ,
lZoe T0 M 9 S rn 0 5
/
4 . ) ev e/fi f3 u t 0 Js i e e1dn _ i t t r/ o _
.t ( e 3wc e 7 t + a 5 e m nm5
- : :: 1 r a s e
,o r eea t el NMTS ,t u l rel s a ei AA C t cT nD cF e OEEVNdmK TNCD E
T E TE ni e yS s c CEay wC o i f L t o i R) c r u RI ON EL LE att c n d0 a2t2 c u n oL rE S o DU PT eA e r1i r of ( s w t t i a e 2 n. ni P . t u o t st st ) m . O a r nt r r a t r2 a . C P eaeU h I r e o a e pO S . 2etg . G p h eT I e sR - , G t and t r t it ei i P yt E i _ W uodrg b r R k i h o n s f EE rt eU M T i 5 t n k <% e n oS A e pe / g . c,
- do n a t
CC u L
.)f O
I 8 g n ~ i t , 5 m s M e A i , t e / ~ D I sd y o M b - T SC mf C a o J r JE E R 9 1 .
.I
- mc n aS o r i R o c ct n h = N S i a o ,
-- E + e A
t ui G -9a st t l 1 a nt P D
' nee e
cv p r / N a a. mic r d:o auna umt ot u e ndh - N iE e o et _
- E h '. M A A
L R O t -J - - s H T N li M e1 23 M34 - - . . eD n 3 OM w o .
-D T T R
fD o gr J n F O o P dn o i S. E' R oi t ht c 1 t e - T I tt a n eue i tm Mh R 5 w _
}l
- x
. - E 7 e s n o
)
n n e
)
n _. s .i .c - a C a i ia , e l L i - R p n o - od i ! .b ( t -tc a n b ( n s o) e e ug r i s rT r n o t r a u 7 u r) r o 4 l r ts in e r n i o ei vE i uM D( 3 4 a F e
.R ai rt l l a
c yT a a p _ t c gr r n n a n o rt
, e r. ' t o si(
ob _ -. I m el at i t
- pt el a n a r e
'S ni n u u ri r e p o uif je meh y I s q a e ep d pt T aE M R R OA OO e - - - - - - -
R A BCDE F GH 2 2 9 e / t 3 ' ' a 1 d . D / e~ _ 3 l d u ed c e rh
. o c
. F S ,
o 2' N : : F 5 _ I I
' (; ;! l :,,liij; ;!! i;a}ii,1!ii :jiT ii3;4< i4it tdi }}! i1! . i41'-
m... . . _ - . . _ _ _ _ . _ _ _ _ - _ _ . _ - . _ _ . _ . . _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ . . _ . - _ _ . . - - OPSEG . ht+c seru e Doc 9 r w Cas company P.O t>..- ;06 Heun Lc Unam Nov Josey 060m Hape Creek Geneatmg Staban I' April 15, 1992 l l l U. S. Nuc1 car Regulatory Commission i Document Control Desk Washington, DC 20555 Deer Sirt i MONTilLY OPERATING REP 0FT HOPE CREEK GENERATION STATION UNIT 1 DOCKET No. 50-354 In compliance with Section 6.9, Reporting Requirements for j the Hope Creek Technical Specifications, the operating statistics for March are being forwarded to you along with the summary of changes, tests, and experiments for March 1992 persuant to the requirements of 10CFR50.59(b) . Since oly yours, i
@t o_ -s
.J Hag n Gener[41 lj nager -
l 1 ( Hope Crc 6k Operations RAR:1d l Attachments C Distribution / l
\
[gf19 !; n 3 n ep : i , s' 1Yli6 EMornv Penn'n I A# 9204240103 920331 . PDR ADDCK 0D000354 4 1,w w : a3 R PDR :
1 r 6 4 . IliDEX 1 14 UMBER SECTION Q U AgIJ ; Averago Daily Unit Power Level. . . . . . . . . . . 1 l l Operating Data Report . . . . . . . . . . . . . . . 2 l Refueling Information . . . . . . . . . . . . . . . 1 ), i Monthly operating Summary . . . . . . . . . . . . . 1 1 l summary of changes, Tests, and Experiments. . . . . 8 J l . I a f 7 i l l-l- l l 1
l
. AVERAGE DAILY UNIT POWER LEVEL ,
DOCKET NO. 50-354 _ , _ UNIT Eqp_q Cregj DATE .4/15/92 i COMPLETED BY V. Zabiolski ' TELEPilONE ..(609) 339-3506 f i MONTil dgrch 1992 ; 1 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWo-Not) (MWo-Not)
- 1. 1Q11 17. 11 j.
, 2. 1062 18. in
- 3. 1057 39. E11 i
- 4. 1054 20. 1045
- 5. 1Q14 21, 1068 6 .- 2 14 22. 1070 '
- 7. Q 23. 1069
- 8. k 24, 1QQ
, 9. A 25, 1063
- 10. A 26. 1Qin
- 11. a 27. 1059 ;
.12. A 2 8 .- 1049 .
l 13. A 29. 1059 -
- 14. A 30, 1Q11 i 15. .9= 31. J060
- 16. 'A l t
I
)
_ . _ _ _ _ _ _ _ . - _ _ _ _ . _ _ _ _ - . . _ ~ _ _ _ - _ _ _ _ - i
, OPERATING DATA REPORT l DOCKET NO. 50-354 '
UNIT LLQp.9 Creek DATE - 4 /15 L9 2. COMPLETED BY V. ZabinhkL3 e 7 TELEPHONE 19.09) 339-3506 t
' OPERATING STATUS
}
- 1. Reporting Period March 1992 Gross Hours in Report Period 214
- 2. Currently Authorized Power Level (MWt) 329_1 Max. Depend. Capacity (MWe-Net) 1.qn Design Electrical Rating (MWe-Net) 1067
- 3. Power Level to which restricted (if any) (MWe-Not) ligng
- 4. Reasons for restriction (if any) ,
This Yr To ' lionth DAt2 CRaulative
- 5. No. of hours reactor was critical 522.8 1962.8 39.124.1 l
- 6. Reactor reserve-shutdown hours .Q ,_q M M
- 7. Hours generator on line- 495.2 111L.2 38.509 1
- 8. Unit reserve shutdown hours M M M
- 9. Gross thermal energy generated 1. S 2 3 1Q) 6.255.121 122.252.263 (MWH) 10 '. Gross electrical energy 510.040 2.0952 580 4 0. 4 4 8_ Q21 generated (MWH)
- 11. Het electrical energy generated 483.62jli L OO2. 9 31 38.654.480
- 12. Reactor service factor 70.3 89.9 84.5 ,
- 13. Reactor availability factor 70.3 3.M E4.4 ,
- 14. Unit service factor jid 3_gL.3 g_L2
- 15. Unit availability factor 66.6 31.5 BL2 16.-Unit capacity factor (using MDC) 63.0 89.0 81.0
- 17. Unit capacity factor 60.9 S.S d 78.3 (Using Design MWe) .
- 18.. Unit forced outage rate M M id !
- 19. Shutdowns scheduled over next 6 months (type, date, & duration):
Refueling outage, 9/12/92, 60 days ,
- 20. If shutdown at end of report period, estimated date of start-up:
N/A. I
-___________________~_.m_..._-__ ... - ._ ._.. _ _ __. ,_ ____,-._ ,,,,,.m m - ,,,_ .,,,._,...,...,,_,.,-,_,,-..-,---r.
i
. '. l'
. OPERATING DATA REPORT UNIT SliUTDOWNS AND POWER REDUCTIO!ls DOCKET NO. 50-354 UNIT llope Cm};
DATE __4/15/92 COMPLETED BY L_InkinimKi_ TELEPl!OliE (609) 339-35Q1 MONTil-March 129,1
]
l MET 110D OF SHUTTING DOWN Tile TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE ! No. DATE S=ScilEDULED (!!OURS) (1) POWER (2) ACTION / C:.+M JNTS 1 3/6 S 248.8 8 1 Scheduled - Maintenance Outago - J i i ; Summary , L
.-~..--,w.~. ..--.-..-.;:-...,,....-.-.-,:,
1 REFUELING I!1FoluiATIOli DOCKET HO. 50-354 UNIT 11@.e Creek DATE _4/_11t/ 92 COMPLETED BY L_ligJ1inggXorth TELEPHONE .1609) 339-1051__ i MONTH liatc.b 1921
- l. Refueling information has chanejed from last monthe Yes No X
- 2. Scheduled date for r$xt refueling: 9/12/92
- 3. Scheduled date for restart following refueling: 11/11/94
- 4. A. Will Technical Specification changes or other license amendments be required? ,
Yes No % B. Has the reload fuel design boon reviewed by the Station i operating Review committee? Yes No X If no, when is it scheduled? not scheduled (on or prior to 7/24/92)
- 5. Scheduled date(s) for submitting proposed licensing actions 11/.A 7
- 6. Important licensing considerations associated with refunling:
-Same fresh fuel as current cycle: no new considerations ,
- 7. Number of Fuel Assemblies:
A. Incore - 764 B. In Spent Fuel Storage (prior to refuel.ing) 760 i C. In Spent Fuel Storage (after refueling) 1008 L
- 8. Present licensed spent fuel storage capacity: 4006 Future spent fuel storage capacity: 4006-9.- Date of last refueling that-can be d.tr charged 11/4, 2010 to spent fuel pool assuming the present (EOC16) licensed capacity:
(does not allow for full-core offload)
.,v ,.-v.w wy ,- y wev r w w vi web-+*,w,-.,.-e----ur-.-+h...,-- ey .,v- ww--.. .-w.-,,-w-. u,+s- mwmr,ew-e---s.----~,s.-y,-..,-_,_mr_,._----.n.--..,m ---m-~-- - .-.
. r HOPE CREEK GENERATING STATION ;
HONTHLY OPERATING
SUMMARY
March 1992 Hope Creek entered the month of March at approximately 100% power. The unit operated until March 7, when it was manually shutdown for a scheduled maintenance outage. The plant completed its 300th day : of continuous power operation prior to the shutdown. The reactor went critical on March 16 and the plant was brought back on-line at 1412 on March 17. The plant operated for the remainder of the month without experiencing any otner shutdowns or reportable power ' reductions. On March 31, the plant completed its 14th day of continuous power operation. I f
+
t F l i t _,a e_,w...,,, --E4=,.e,w,.-. ,. ,_ ..,,.-c....,, ,.c-,ww.,,_,m,-. -,-- e.,._4-y.,m . . .,,3.r e
SUMMARY
OF CilAliGES, TESTB, AllD EXPERIMEliTS FOR Tite 110PE CREEK GEllERATING STATIOli MARCli 1992 E' 1 I l I 4 h
,-,ey ,_- .-.am...-. .. , m,. . . _, ..,,,,..y,. ,, , . . . , . . . _ . . . . , . _ . , .-,,...mm, . , . , ... ,, ...e..,w_.g . . , . . , ,,.y., .-._,,,.,m.m.,,..
- - . _ - _ . . - - . . . ~ . . . . . _- - - . . ~ . - . - . . . -
Th's following items 1. ave been evaluated to determines l Ic If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or i
- 2. If a possibility for an accidr.nt or malfunction of a different type than any evaluated previo,taly in the safet analysis report nay be created; or i
- 3. If the margin of safety as defined in the basia for any technical specification is reduced. ;
The 10CFR50.59 Safety Evaluations showed that these items did not < create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor. These items did not change the plant effluent releases and did not alter the existing environmental impact. The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved. 4 k l i-I -4 _ . ,. _ _ _ _ _ _ ; _ . _ ._. _. _ . _ _ _ _ _ . . _ . _ _ ,. .
DC' Description of Safety Evaluation 4EC-3112/13 This DCp replaced motor operated butterfly valves in the ' B' service Water Pump and Strainer. They were replaced with upgraded butterfly valves that are designed with metal seats for extended wear and minimal-maintenance. The internal control logic of the operators has been revised to incorporate torque seated valve operation instead of limit switch seated valve operation. No'Unreviewed Safety Questions were involved because the valve replacement and modifications do not adversely affect the safety-related function of the Service Water piping. Additionally, the use of a metal seated valve will provide improved shutoff and performance resulting in minimizing the potential for strainer basket damaga due to valve seat failure. 4EC-3114/06 This DCp replaced Service Water Strainer Backwash ! System spool steel piping. pieces with 6% molybdenum stainlessAdditionally, a butte be replaced with an upgraded butterfly valve that is designed with metal seats for extended wear and minimal maintenance. The internal control logic of the operator has been revised to incorporate torque seated valve operation instead of limit switch seated valve operation. . No Unreviewed Safety Questions were involved because the spool and valve replacement and modifications dr not adversely affect the safety-relatedfunctionoftheServiceWaterpipin!p.ing Additionally, the use of the new material p and a metal seated valve will provide improved per#ormance resulting in minimal maintenance. 4EC-3114/08 This DCP replaced Service Water Strainer Backwash System spool pieces with-6% molybdenum stainless steel piping. Additionally, a buttorfly valve will ; be replaced with an upgraded butterfly valve that is designed with metal seats for extended wear and minimal maintenance. The-internal control logic of the operator has been revised to incorporate torque seated valve operation instead.of limit switch seated valve operation. No Unreviewed Safety Questions were involved because the spool and valve-replacement and modifications do not adversely affect the safety-related function of the Service Water piping. Additionally, the use of the new-material piping , and a metal seated valve will provide improved performance resulting in minimal maintenance. I e,o J...- . , , , , n- .--,--e-m , - , . , . ..-r - --,,m .,,.,.,-.....,-,--,.,.-.-,,-.,.-+,,-.~m.
'. l DEE Descriotion of Safety Evaluation 4EC-3226 This DCP modified the logic of the 'E' and 'F' Filtration Recirculation, and Ventilation System Recirculatkon Fans. A permissive interlock was added to the low flow automatic start signal. The permissive is automatically activated by the automatic start signals for the Filtration, Recirculation, and Ventilation System train. The permissive is reset when the 'A', 'B', 'C' and 'D' fans are stopped and the HI RAD or LOCA sig,nals are no longer present.
No Unreviewed Safety Questions were involved because all of the emergency starting circuits remain as originally designed and the design does not change the number of Filtration, Recirculation, and Ventilation System Recirculation Fans available for normal or emergency operation. . 1 4EC-3326 This DCP replaced the ring-lug termination of the ! servo-valve and position transducer wiring of valves ir. the Turbine Generator System. This provides additional support at the crimp point of the connections. This DCP did not change any of the operating parameters or the controlling fu :tions of the equipment. The sole .o provide a more reliable termination; purpose was no Unreviewed therefore, Safety Questions were involved with this DCP. 4HC-0212/05 This DCP installed a 600 gpm mechanical side-stream filtering unit with a 40 paid differential pressure monitor and all of the associated piping, fittings, valves, and instrumentation. The filter will cleanse the Turbine Building Chilled Water system of insolubla solid impurities and lubrication oil that hamper: system performance and reduces the sa>vice life of the equipment. The Turbine Building Chilled Water system has no safety related functions with the exception of the isolation valves located at the-Drywell penetrations.--The proposed modification has no impact to the main flow to the pumps and the design includes a bypass back to the suction huTder of the pumps in the event that the filter is down. Therefore, no Unreviewed Safety Questions were involved with this DCP. l
l l l IMR Descriotion of Safety Evaluation ; 92-001 This TMR installed electrical jumpers across the l Feedwater Henter's High High Level Trip Switches. ' These switches cause spurious high level trip signals during low power levels due to inleakage in the reference leg. The jumpers were removed , after the level signals stabilized. The Feedwater system is not safety related and is ' not required to be operable following a LOCA, other than for containuent isolation. Failure of the Feedwater system does not compromise any safety related system or components. This TMR has no impact on the containment isolation function of the Feedwater system. 92-003 This TMR removed the overload heaters from the breakers for the Reactor Water Cleanup Discharge to l Condenser Valve and the Reactor Water Cleanup Discharge to Equipment Drain Valve. Removing the overload heaters from the breakers will prevent the valves from inadvertently opening during an Appendix R fire. Disabling these valves, along with the overhead annunciator, does not prevent their associated systems from performing their designed functions. Also, the UFSAR discusses the Appendix R requirement that the valves be disabled. l l l
. ~ . _ . , _ - _ _ . _ . . _ ,._ _ _ _ _ . _ , _ . . . _ . , _ _ . . _ . . - . _ - - - _ - - - _ _ . - _ - - . - . -
IME Description of Safety Evaluation 92-006 This THR eliminated the possibility of a Residual Heat Removal Shutdown Cooling isolation due to a loss of the 'A' Reactor Protection System power during an outage with the plant in cold chutdown. The TMR climinated all automatic isolation signals to the Shutdown Cooling Suction Inboard Isolation valve powered by the 'A' Reactor Protection System. It also removed one of two Reactor High Pressure interlock tignals (powered from the 'A' Reactor Protection System) from automatically closing the Shutdown Cooling outboard Suction Valve, the two Shutdown Cooling Injection Valves, and the Head Spray Outboard Isolation Valvo. The TMR was removed prior to plant startup. The isolation of the Residual Heat Removal Shutdown Cooling Suction and Injection Lines is provided by motor operated gate valves that are interlocked closed by a kcactor High Pressure signal during normal operation and are automatically closed during an accident by Low Water Level isolation signals. The removal of automatic controls from these valves did not create the possibility of the Shutdown Cooling lines not isolating. Altnough part of the automatic isolation capability is removed, manual capability remains. The solation logic channels are not required to be operable in cold shutdown per the Technical Specifications. Therefore, no Unreviewed Safety Questions are associated with this TMR. t
O Procedure ROY_liiDD DERgriotion of Saf ety I'va19AtiRD llc.IC-LC.AE-0005(Q) This procedure performs a loop calibration Rev 0 of the feedwater flow transmitters to t;. feedwater flow computer points. Individual components are adjusted as required if the loop is out of tolerance. This procedure installs jumpers to bypan: the 20% total feedwater flow interlock La the recirculation pump speed limiter to preclude an actual recirculation runback from occurring during the transmitter calibration. The use of jumpers to bypass the recirculation pump speed runback in the only item in the procedure that in a change as described in the SAR. The recirculation runback logic does not make a significant contribution to the mitigation of the accident and it does not challenge the core thermal margins or vessel preasure boundary before the scram. Therefore, there are no Unreviewed Safetv Questions associated with this new procedure. IIC.SS-IS.ZZ-0010(Q) This procedure provides methods to test l Rev 0 valves to ensure pressure isolation valvo leakage rates, primary containment leakage races, and individual valve leakege rates are met. This procedure changes the facility by installing jumpora and lifting leads to defeat the liigh Pressure Coolant injection and Reactor Core Isolation Cooling Low Steam Supply Pressure signals. The procedure also rotates a Breathing Air Spectacle flange, replaces the Primary Containment Instrument Gas Drywell Intake Screen with a test flange, and installs a test plug J n the liigh Pressure Coolant Injection Suppression Pool. The use of this procedure does not increase the potential for draining the Reactor because the Prerequisites require that the Main Steam Line Plugs are installed or that the Reactor Vessel level is maintained below the Main Stear. Line Nozzles. Also, adeq.uate precautions are taken to prevent equipment malfunction during the performance of this procedure; therefore, no Unreviewed Safety Questions are involved.
9 UFSAR Section Description of Safety Evaluation ; Table 11.5-1 During the Radiation Monitoring System I Table 12.3-9 Configuration Baseline Document review, l several typographical and editorial deficiencies were identified. One of the resolutions involves updatina the UFSAR to reflect the correct information. The extent of these changes are correcting editorial and typographical errors. Therefore,ith involved w this UFSAR change.no Unreviewed Safety Questio { r-I
--._ .-,...-... . . . . - , _ - , - . , . . _ . . . , - , - . . _ ,}}