ML20092F757
| ML20092F757 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/01/1989 |
| From: | Fitzwater L GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20092F288 | List:
|
| References | |
| CON-IIT05-002-131B-90, CON-IIT5-2-131B-90, RTR-NUREG-1410 LO-HO-60315-001, LO-HO-60315-1, NUDOCS 9202190490 | |
| Download: ML20092F757 (20) | |
Text
USERS.00P_Y GEORGIA POWER POWER GENERATION DEPARTMENT V0GTI.E ELECTRIC GENERATING PLANT
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TRAINING STUDENT HANDOUT TITLE:
LOSS OF RESIDUAL HEAT REMOVAL NUMBER:
LO-HO-60315-001-02 PROGRAM:
LICENSED OPERATOR TRAINING REVISION:
2 i
AUTHOR:
L. FITZWATER DATE:
8/9/89 APPROVED:
I l
DATE:
b
,O..
9 l/,/ k 1 VEGP PROCEDURE 18019-C. REV 5 a
SOER 88,003. LOSSES OF RHR WITH REDUCED VESSEL WATER LEVEL AT PWRs 1
9202190490 920116 PDR ADOCK 05000424 M
'I 5
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caer vroo3 ON OC108EQ 19TH AND 20TH, 198R, SOER 08-3
- LOSSES OF RESIOUAL HEAT REMOVAL WITH RE00CEO REACTOR VESSEL WA?EQ LEVEL AT PWRa,"
WAS MATLEO TO ALL INPO MEMBERS AND PARTTCIPANTS.
THIS SOER IS SFING TRANSMJTTEO ON NETWORK FOR MEMBERS AND PARTICIPANTS THAT USE ELECTRONTC MEANS FOR OfSTRIBUTTON.
SIGNIFICANT OPFRATING EXPERIENCE REPORT:
88-1 YELLLOW PROMPT ATYENTION October 19, 1988 IO33ES OF RESIOUAL HEAT REMOVAL WITH REQUCEO VE3SEL WATER LEVEt AT PWRs EVENTS:
UNIT (TYPE):
OTABLO CANYON 2 (PWR)
WATERFORO 3 (PWR)
OOC NO/LER NO: 50-323/87005 50-382/86015 FVFNT DATE:
4/10/87 7/14/86, 5/12/88 N333/AE:
HESTINGHOUSE/ PACIFIC COMBUSTION GAS AND ELECTRIC ENGINEERING /EHASCO UNIT (TYDE):
3AN ONOFRE 2 (PWR)
IION 2 (PWR)
)
OOC NO/LER NO: 50-361/86007 50-304/85028 EVENT DATE:
3/26/06 12/14/BG NSSS/AE:
COMBUSTION ENGINEERING /
BECHTEL SARGENT & LUNDY l
RFFFRENCE3:
1.
INPO Significant Event Report (SER) 15-97, " Extended Loun of Routdual Heat Removal OurIng Steam Genarotor Maintwn6nce" 2.
IN90 Signfficant Event Rwport (SER) 35-86, " Extended Loss of Shutdown Cooling Due to Steam Binding of Shundown Cooling Pumps
- 3.
IN90 S'fgni ficant Event Report (SER) 17-86, " Loss of Shutdown Coolfng Flow' 4.
INPO Significant Ops.r a t i m; F =,.* r l a n c e Report (SOER) 85-4,
" Loss or Degradation of R.u ' d u a l Huat Removal Capability In oWRs*
5.
INPO S igni f icant Event Quom r (SFR) 23-86,
- Loss of Decay Huat 9amoval Flow Due to !~.Nouate hactor Coolant Sys t ee-
r
-Level control" S.
INPO Significant Ev en't Raport (SER) 31-86,
- Loss of Residual Heat Removal Flow Due en' inadvertent Oraining of the Reactor-Coolant Sys t ern" 7,
TNPO Significant Event Raport (SER) 2-87, "Onge*adation of
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Erneegency Core Cool ing" 8.
INPO Operations and Maintenance Reminder (O&MR) 295, " toss.
of Realdual Heat Removal Flow"
- I 9.
Nuclear Safety Analysis Center Report PSAC/52 of-January __
- 1983,
" Residual Hunt Removal Expuriance Review and Safety-Analysis, Pressurized Wwter Reactors
- 10.
NRC information Notice ~88-36, "Pousible-Sudden Loss of RCS inventory Ourin0 t.ow Coolant Level Operation," June 8, 1988 11, NRC informntion Notice 06-101 " Loss'of Decay Heat Removel i
Due to Loss of Fluid Levels-in Anaccon Coolont System,"
December 12, 1986 12.
NUREG-1269, " Loss of Residual Hest Removal System," Diablo Canyon, Unit 2, April 10, 1997
-1 13.
NRC Letter of May 18, 1987, ' Loss of Oscay Heat Removal Function at Pressurized. Water Reactors W*lth Par $tially 3
Drained Reactor Coulant' Systems" 14, NRC Information Notice 87-23, ' Loss of Decay Heat-Reinovel OucIng Lnw Reactor-Coolant Level Operation," May'27,_1987 15.
NRC Generic Latter 87-12
" Loss of Residual-EHeat Ramoval L
(Rha) While the Reactor Coolant Syarem (RC3)
(
Filled,* July 9, 1987-
' is: Partially' 16, Presuntation paper-entitled 'An-Improved Reactor Coolant-Level Monitoring.3ystem to Prevent Loss of Residual Heat I
- Removal Function-in a PWR " 13th Bien'nial Conference on Reactor Operating Experience, August 30 - September 3, 3
1987-' Chicago, Illinoim, by Dina.M., Lawrence,_. Commonwealth.
-Edison:
NOTE:
The following terrns are used: In.tcrchangeably wIthin this-SOER:
shutdown: cooling-o o
residual heat vernovel o
necsy heat removal-
- o. low pressure safeey injectfun 1
SUMMARY
i a
4 c....
(.,
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SignifIcant Operating Exp-rienco Report (30ER) 35-4,
- Loss or I
Degradatfon of Resfdual Heat Rumaval Capability in PWRs,"
l issued in Augus t-19 8 5, discussed events involving loss or l.
degradatlon'of residual heat emnovel (RHR) capabil-fty at pressurtzed water reactors.
The events focuswd on tho thr'eu veuw t comnion ways of losing RHR:
o low cenctor vessel lavs1 resulting in loss of RHR pun.p xuc t inn o
clogue of t;m RHR pump suction valva telppirig of the runn ing RHR purup l.
o l
F ighty per coi.t of the staticns have i mp1.ernan t ed roos t of the l
o ur. orm n a n d o t i o n s con ta inwd in SOER 85-4, und approa.-tmaculy 40 t
percent ha v,e t riip 1 wm.+n t e d
,e l l the recommendations.
T he nios t fiequent r e c urnrn e n d a t i o n s awaiting 1 rnp 1 wenun t a t i on aru l
Independan t reactor vassal level indication and soros procedure l
changus.
l Howwvue, events invv1ving loss of rew'idual heat e arnove l capabflity with reduced reactor vessel
- u. ster level continue to high rate.
More than 10 evants involving the loss occur at a l
of residual heat veraoval capabi'ity f o r' g e a a t e r' than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> have occurred since August 1985.
Thlw SOER discusses fIvo a*
these events, including three wFich resuited in boiling of wa t e r' in the reactor core.
Thes6. events could hava been preventud through a thorough t'aview and effective end timely irnplemen ta tion of the recommendations in SOER 85-4.
Thesu-events d ernons t r a te the need for increased attention to activitius that require opet'ation of the realdval heet-remov,sl system with reduced reactoW vessel water level at mid-loop.
This SOER suppluments 30ER 85-4
" Loss.or Degradation of i
Resit.'ual Heat Rernoval Capability fa PWRs,* and providas additional recommendations to present a loss of core-cooling l
with coactor vessel water level at mid-lonp.
These events are significent because loos of residual' heat v ernov e l capability can result in tha boiling of cooling water, w ft.h the potential for core uncovery and damage.
The losw of residual heat t'emoval cooling can also lea.d to' airborne radioactivity releas,es, Increased radiation levels due to loss of core shielding, and equipenent damage.
OFSCRipYTON:
DI A/JLO CANYON 2 (4/10/87):
The plant hed been shut down for seven days.
The.' residual hunt remova l sys tern was in operation.
The reactor coolant system had bean-drainnd tee the mid-loop lwvel to permit. rernoval 'o f the steam generator pcimacy side manways for norxlia - darn Install,etion.
The control room ope +rator was monitor'ing: the-reactor vessel water-level usIng a temporary-water level indIcetion syatom.
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i An enginner prepartog for a local leak rate teat on the naactor coolant punip a nta l return line to thw volume control tank (VCT) opened a vent and drain valve to drain the water from a previously 1'solated lina.
The engineer then left thw l
containment.
While the line weas dra in ing, one of the boundary l
holatinn valvos leakad.
This resulted in a loss of inventory from the VCT.
Thee dra in path was ont rnonitored locally for continued lakkage.
The opar,etors were unawaru that a drain l
path had bwen established because the enginuwe propiring for j
thw tust uns working under one-day-old clearance eaquest,a n d u
l had not befafed the on-duty opnaating s;eift before starting this uvalutfun.
1 l
The rusultIng c! rop in the VCT 1evel was observred f rumad ia t e 7 y in-
{
the control ecom.
Reliev Ing the drop in VCT l e v w 'l was dua to i
dacreased lwt down flow the operators increased lacdown flow to taatntain constant VCT lovel, as s uin in g that this action would alsc rooin t a in enactor vouse) water level constant at in i d-l uo p.
As - a ruuit,,setual rwactoe sessel-water level began to deureasu slowly as indicated on the temporary water level ind! cation system.
The increased letdown flow had lowered t
reactor vesse'l water level until air on t ra inneen t occurrud through a vortex at the residual heat removal-suction nozzle connection to the reaccor coolant system hot-leg piping.
This resulted in air b-ind ing ano cavitat ion of both the operating and the ntandby residual heat runoval punsps.
When the decreasing reactor vessel water levol was detacted, tha operators isolated the letdown and charging flow paths.
Sncuring letdown flow stopped the loss of inventory from the ewactor coolant system; however, becausa the open drain valve had not been detected, the level in the VCT continued to l
decrease.
1 i
When the residual heat removal pump cavitated, residual heat removal cooling capability was lost,.and the reactor coolant L
t enipara ture -bagan to rise dee to decay heat.
The, operators underestimated the heatup ratu because cha ir - previous _.
empartence with a loss of residual heac eamova~l.accurred when the coes had little decay heat.
Because the core exit thermocouples'had been disconnectud in preparation for reactor closure head comoval and the-reactor cuolant loop resistance-twmperature detector indfcations were unreliable due to lack-of i-reactor coolant flow, the-operators were unaware of the capid temperature increase, Conceras for the_ safety of personnel removing the steam generacce manways resu;ted in delays in ewising reactor vessel water-lavul.
As a result, reatoration of reactor vessel water level was delayed for' one hour and 29 minuten.
Reactor coolant w ys t am n c m comparature increased-from=87 degrees to 220 degesws Tahrenheit, 6nd reactor coolant system pressure increased f roru attnouphoric to approximately-10 psig.
During this period, steam was vented to containment via thx reactor vauel head - teruporary _ vent line that ruptured dem to tho,orassure increase.
Water was spilled from the partially la e
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unsealed senam generatcc roanways.
/1cborne radioactivity l
1evels in con': ainruen t rose, requir-ing evccuatton of personnel The plant has revised its procedure for draining the reactor
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cool nt system, providing precautions relat;ng to residual heat t*moval flow and reactor coolant systuw level to precludi afgnificant air entrainment due to vorte w formation.
A ptwowquisits that requires effective corumunications during all work ac t-f v i t f es that dieuctly or indfructly affect reac t or i
conlant system inventory also was included.
The loss of e em f dua l hea t-ruinnval flow at nor rnal procedure how also been raviswd to recognize that if en operating residuel he s c removal pump cav. ta tus excessively or losus wucelon, the Idle pump should not be started until adequato reactor vwssni water i ome l is restored.
In addition, the a bnortua l procedure now l
includes a table for d e t e rrrin i ng the time (based on
- e. wisting water inventory, power history, and time since shutdown) unit i l the reactor coolant system temperature will reach 200 degrens Fahrenheit w thout forced cooling flow.
- Finally, i
t contafoment integefty is now required to be established prior to mid-loop uporation.
Additional datafim of this event are provided in Referencn 1 WATERl*ORD UNTT 3 (7/14/86):
i l
The plant was shut down with the shutdown nooling syJtem in operation.
The reactor vessel water. level was being monitored locally by flexible vinyl tubing and in the control roo:a by - tha l
n'eaccor vessel water' level. monitoring system.
T h e - o nco a a t o r' s began draining the reactor coolant system to the mid-loop level l
to replace a reactor coolant pump seal.
Two drain paths worn betag used to lower ruector vessel water-level. One path'wac the nor mal drafn path and was addressed in-thu procedure.
Thu second path, which was not addressed,in the procedure, was used to speed-the dra-Ining and to conserve water by draining.co the refuel?ng =ater storage pool.
When the desired reactor vessel water level was reached, the first dra1n path was secured in accordance W ith the procedure.
However, the second path was I
overlooked, and water continund to drain Ecom the coactor coolant system.
l l
The operators were not aware of the continuing draining becwuae a vacuu.a in the reactor coolant system had collapsed the l
-flexible vinyl tubing used for.luve' indication,_ causing an inaccurate reading.
In addition, the reactor vessel water l
lavel monitoring system in the. control room indicated only large changes in level (marked fn 4 foot increments) and thwrefore. did not have the accuracy necessary to monitor and-maineain level at mid-loop.
Whon the water level had decreas9d nuffictantly, thw low presvura e,ofuty inject ton pumps s: tart ed -
cavitating from air' en t r 7 i nraan t.
The operators then identiffed and secured the second drain path.
15 4
_m Restoratton of shutd awn. cooling was delayed for i 1/4 hours while the mechan ics wor'hing on tha r eac tor' coolant pump seals wer e evacuated fr'om contaihtcent, and repeated attompts were made to vent'the low pressurw ufsty injection purn pa.
The low reactor veesel water level'haJ caused a staam bubble to form between the shutdown cool f rig ifne loop uaal and the pump i, uc t ion.
The steam could not be condensed due to aaturation
+
conditions and could not be ventec, dua to the limited capac-fty of the vacuum pelruing system.
The t'eactor coolant system water temperatura inct'ensed f rorn 138 to 232 degrees Fahr'enheit i
(., a t u c a t i on ) bwfore uhutdown cooling flow wau reestabitshed.
During this t f rua, boiling occur'ed in the core ragion. Shutdown cooling was restored by Jogging the inw pres =ura safety injectton uurnps and t'autr'culating cooler wa t e r' from downstr'eam of the shutdown cooling heat exchar wes to the pump suctfon.
in this ena nn e r, S t uarn in the pump suction-linea was cooled and condensud, n11owing oper ation of the pumps without cavitation.
The lack of guidance for t'estorIng shutdown cooling contributud to the difficulty in s'ecovering from this condftfon.
The plant revised the procedur es f o r' draining the reactor coolant s ys t aru to prohibit uwing ruul tiple dra in paths while the pressurizer
- is empty and to provide details for e as tor' Ing shutdown c ool 'i ng.
Additional details of this event are provided in Re f er'ence 2.
s l
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WATERFor'O UNTT 3 ( S/12/S 8 ) -.
1 The plant was shut down-with che shutdown cooling system in operatton.
Rea c to r' vessel water level was befog lowered to removu the s tearn gener'ator nor;: 19 da,ms and to test a new digital reactor vessel water level indicator.
A flexible viny' tube was used to locally monitor reactor-vemac? water l ove.l.
Th6 dealning evolution was socured when inconsistencies dw eloped between_the dfgital weter level ind(cator and the flodble vinyl tube.
A f ta#> wa t er' was removed from the nnemally dry rwference leg of~the digita's level iridi ca tor, the digital level indicator-and flexible viny 1' tube were brought back into apparent a g r' a emen t.
The opscators then reeuwert draining the reactor vessel.
At an indicated level of appr ru irn6tely 10 feet on the flexible v inyl tube and 14 fece on the a*gital Instrument; the
- A*
low pressure safety 1njection purup bwgen to cavitate and was secured.
The operator's ra ised r'wactor vessel water leve's and stmetud the
- B' low prensure sofsty injection pump in accordar.ca with thn obnormal procedure, limiting-the t'eactor coulent -ystem water t eropera t u m, e i,e to a few degrees.
l.ow posssura safety-Injection p o '"o "A" was veinted and return 6J to servfce.
After plant pursuon I
'n,pec t.ed t he installation of the two level ind ica tor's, thw
+=rators t'ecommenced dr'aining the reacto" vensel.
However.
+
,,.: ction did not detect a lo
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loop seal in the flonible vinyl tubo.
I inaccurecies fry the new digital Because of previous indicator, reactor vessel water level for measuring reactorthe aparators. relied primarily on tube the flexible vinyl vessel watea level.
Oraining of the reactor vessel was secured when the fler1ble vinyl tube indicated a level of 17 fast.
-(The digital Instrument i
indicated 13 feat.)
Thirtean feet ha below the centerline i
(13.38 feet) of the rwactor coolant i
loop hot-leg piping from which the low pressure inject 1on pumps take auction. _ Shortly after the draindown was secured, low
- A" pressure safety injection pump again began to cavitate.
The reactor vassn1 water lavel was raised, and shutdown cooling was proioptly restored.
A datofled invest igation of the Ipvol indication problems found 1
that thn flexible vinyl tube had been rerouted to allow installation of the new dfgital for level indicator.
When the flexible vinyl tube was rerouted, a loop seal went undetected and caused reactor vessel was created thst indicate high.
water level ta The pl an t revised applicable p"ocedures tubing length.s in the reactor vessel levelto minimire vinyl In addition, volume of waterprocedures~are being changed indication system.
to identify the in the reactor coolant system for specific ledfcatwd reactor vessel water level heights (with and without steam generator nozzle dams in place).
of the water level indications by correlating the amountThis providas a check water drained from-or added to the of the indicated water leve l change.
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SAN ONOFRE 2 (3/?6/86):
Tha plant haad installed andin a refueling outage with the reactor vessel was The heated-Junction thermocouplesthe shutdown cooling system in operation
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level for the reactor vessel water fndicating sgatem and been removed in preparation forthe core exit thermocouples had reactor coolant refueling.
Water level installed cwfueling watersystem was being - moni tored by a recen tly d n thu level provided both narrow-and wide-rangeindication system The. system foe reactor coolant systemloop resistance teaperature detector room.
The used being temperature indication.
Rather than relying molely upon a local flex 1ble vinyl tube indicator, as had been done in indication systern previous outagasi the refueling water level l
Increased accuracy and operator installed to'orovide was refilling evolutions.
control over draining and In preparation for nepairing a leeking steam generator nozzle dam, the operators began lowaring and placed the reactor coolant reactor vessel water level afeborne contaminationsystem educcon, which minimizas' containment system integrity is breached, in meev ice.
Both of the control when the reactor coolant 17 s
w w
room narrow-and wido-rango refueling wotor lovol indicators began oscillating after the-eductor was placed in service.
When plant percornel were unable to careect the oscillation problema, the flexible vinyl tube indicator was installed.
The opwretors, who distrustwd the refueling water level indication system due to the oscillations, continued draining the reactor coolant system, relying primarily on the flexible vinyl tube for level indIcatfon.
Aftwe a ppro m inia t w l y 2 1/2 minutes of draining with no apparwnt level chang *, the shutdown coolinn system flow to the reactor coulant system was decreased from 3,000 to 2,000 gallons our m inu ta to divert more drafn flow to the refueling water storage t ank and to reduce thw potential for pump vortexing.
With ein indicated level of minus 78 inches _on the floatble vinyl tubo, ewactor coolant system draining was stopped to verify lavs1 and stabilfie cooling flow.
Initially, the $butdown cooling system showed no sign af a ir entrainment, and the indicated level was 5 inches above-the minfmum level allowed by proceduro (minus 78 inchws is equivalent to 1 1/2 inches above the hot-leg pfping midpoint).
A mhort time later, large motor current oscillations were observed on the operating low pressure safety injection pump, and the purop was stopped.
Because the flexible vinyl tubing indicated adequate level, the low pressure safety injection pump was restarted.
- fter several minutes of stable operation, motor current oscillations recurred, and tha pump was-stopped.
The other low pressure safety injection pump was started, and stable operation was observed for several minutes'before the I
suc t ion pressure dropped to zero, and all shutdown cooling flow was lost.
The abnormal operating procedure for loss of-shutdown cooling was initiated, and an accelerated system venting scheme was employed to restore shutdown cooling system
- flow, invwstigation revealed that, while installing and filling the floxible vinyl tube indicator, ma air bubble-was trapped in the tubing, result!ag-in an inaccura':a high reading (plus 10 1/2' inches).
In addition. the reference scale for the. tub'ing was displaced by 2 1/2 inches in the high d irection, creating a L
total inaccuracy of_plus 13 inches.
Vortex air entrainment L
occuersd when the o;:erators used'the inaccurate ficxible vinyl l
tube indication to lower tha reactor vessel water level.to an Indica t ed lesel of minus 78 inchos (actual level-was minus 91 inches, 8 inches lower than the minimum level of minus 83 inches allowed by procedure).
l Subsequwnt utility analysis of this-event.showed-that the enactor cuolan t_ hot-l eg - t oropera t ure increased from'114 to 210 degress Fahrenheit in abou : 49 minut.6s, with local boil ing _ In the core region.
The: calculated bulk reactor. coolant teroperature did not exceed 200 dogrees Fahrenheit.
Steam and approxirna t ely 2 - curies of radioactivity were released to the con t a inm*+n t via che incore~ flux detantor nozzles in the reactor 18
-., - =
vossel head.
The plant has revised app 1 tcable procedures-for draining thw reactor' coolant system to require oper* ability and use of d t vers i fied level -ind Ica t Ions #end has providod detailud guidance for installing the temporary flex 1ble vinyl tube and refuwling water level Indication systema.
l Addiefonal datafis of thtm evwne are provided in Reference 3.
l Tf0N 2 (12/14/85):
l Thw plant wan shutdown wIth the coactor vesual head installwd, but not tensioned, and the reactor coolant system vented to l
a t tuns p h e r e.
A few days after lowering the reactor coolant level to rnpain an isolation valve, the 28 residual heat remove.1 pump became afebound.
After the operator eripped the residual boat removal putap,. the reac tor vessel water' level indication in the control room.bocame arcatfc.and then pegged h igh.
The operacons incorrectly-concluded that the residual heat removal pump or motor had failed based un no flow indicatfan and-the pump low-current readings.
The 2A rouidual heat removal pump was started to custore cooling-but soon developed abnormal cuterent and flow --ind f ea tiona.. The opera t: ore -
then realized that the reactor vessel water level Indicatlon was in accor and that the residual heat comoval pumps had become einbound due to the low water level.
Approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 15 minutos was required to restore residual heat removal cooling.
During this time, reactor coo'lant system water temperature Increased 15 degrees Fahrenheit.
8ecause of repeated problems controlling reactor vwasel w ter level during mid-loop operations, the plant performed a l
detailed rwview of the reactor vessel water level indicating l
- Systems. It was concluded _that the accuracy lof the reactor-vessel water level indication couldche increased by design improvements, The 10-inch residual heat. removal system discharge piping is connected to the top of one of the reacton coolant system cold legs.
The water level sensing l ine for the reactor.vussel level indicating system (refueling waterflowel transmitter) 1 1/2 inch line-connected.to the same reactor coolent wnters a g
sys tem cold log. - _ Both nozzles are in the same vertical plane with the 1 1/2-inch norrle located at 90 degrees with respect to the 10-inch nozzle, when the cold leg was partially filled and a residual heat removal purup was operating.. water froer the-residual heat removal discharge piping ' Impinged on the water surface ulose to the nozzle of the-1 1/2-inch 1.ine.
Because of-the dynamic effects of this impingement, the indicated water.
level was arratic, _especially when reactor vessel water-level was low.
When the water level in the reactor.vossal was below thw nominal mid-point of the cold' leg. the refueling water level transmitter would indfcat-orroneously that_the water 19 w
m
-r-,,
w r
e
4 level was at the mid-point of the reactor coolant system hot log.
The design of Zion's floaible vinyl tubing that connected to the rwfueling weter l o v.41 instrum.*ne_also aff*cted onacatfonal nocuracy.
For this reason, thw v f es ua l readings on the flexible vinyl tube in containment. and the refueling water luval i
t e,nstn f t ter enadouts in the cont rul - ecorn were often not in og egaenen t.
These cuntli t inna mad.a it difficult fer operators to dat *rrn Ine the correct level.
i To provide dependable lwvel indication, modi f ica t ions were made to both reactor vesmal water level. indication systens as j
described in Attachment 1.
l l
SIGNIFICANCE:
1 L
Losu of res idual heat removal capability with low reactor l
vessel water level is significant 1because it can result -I n rapid huatup and boi1-off of the'ramctor coolmo.: and eventual I
core damage.
Wichin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following_ shutdown from a h i g h-pown e aparating history. a loss of residual heat removal-l at mid-loop operation could resul t in boiling in'the core in about 10 minutes.
Even whwn power history and' shutdown time are lens reatrictive, the losslof residual heat ~ramoval. cooling-nt mid-loop could result in-boiling in the core within 20 toLSO minutes wi ti. -subsequen t core uncovering,' depending on tho boil-off rate.
After the onset of bofling, the core could become completely l
uncovered in inss than 10 minutes -i f a' _large cold-leg ' opening (such as en open steam generator ~.nanway).-exis ts. without an adequate reactor vessel vent pa r.h' (such - as the case with the-reactor vessel head ins talled and s coain : genera tor - nozzle dams in place or loop fso13 tion valves: closed).. -In this~
configuration, pressure can gbuild up ' inc the reactor vessel-and-force coolant out of the core'through the cold legfopening.
ANALYSIS /DTSCUSSION:
During-outages, it is frequently necesaary.to_ reduce the reactor coolant level to the loop _ nozzle elevation for l
inn intenance on modifications.
Such operat-fonsJoften cequire-L ma incaining_.the ceactor vessel wa ter level;within very:
costrictive limits.- Under these conditions,-reactor vessel-water level. Increases can result in_coolanc; overflowing through wystem-openings, contaminating.aquipment and personnel, scalding wurkers -and damaging equipment. : Reactor vessel water levul decreases ofLa few _inchus threaten coreLcooling by' disab1 tog the residual heat + ernova l 'sys t em through a i r entrainment.
Because-of the reduced coolant inventory at thase-
- levels, 4-loss'of residual heat ramoval-'can result: 1n rapid heatup rates, boil-of f of coactor coolant, and_ eventual core.
damage; NSAC/52,
- Residual 1 Heat Removal Experience ReviewLand=
20
~.
_v
.i Safety Analysis, pressurized Water Reactors,d.(Reference 9),
notes that after,a loss of. cesfdual huat reinova l cooling,
increases of.*over 100 degrees Fahrenheit have occurred in as 11ttia em 20 minutes in partially drained PWRm.
Because of these concerns, the t f rom spent with the reactor ves.sp1 drained to the mid-loop level abould be minimized when irr'adiated fuel is present, This requires proper planning ar run i n t en a nc e activities, including havIng the necesirany paets, wquipment. per* nnnel, and proceducias available befne a
.mtablinhing mid-locp aperntfun.
H o ws9 v esi, sines mid-loop operatinn cannot be eliminated, addit-ional attention is ownded t a ensure aduquo t e monitnring, cont"ol, and the capability for restoration of reactor vessel wa t er inventory and c we cool ing as described further below.
Loss of Coolan* Inventory:
The timew foe bofling, core uncovery, and ieactor coolant pressurization to occur following a loss of residual heat enruoval are d e p end w. t on reactor power history, t iroe s ince shutdown, and reactor coolant system conftguration.
The tecst restrictive condition occurs for mid-loop operation 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> after shutdown f ollowir.g higr -power operation and when reactor coolant temperatura is above 140 degrees Fahrenhaft.
In this condicion,
') oil ing c.an occur in about 10 minutes after loss of core cooling.
Even for more typ'cally encountered conditions when reactor
- vessel level fa lowered after a few days, boiling could still occur in 20 to 60 a.ifnutes.
Abnormal reactor coolant systeis con *fgurations can result in rapid losc-of-liquid coolant inventory beyond that expected from boiling.
This configuration exists when che reacror coolant hot ~ legs are isolated (e.g.,
steam-generator nozzle daris Installed o.* loop isolation valvow closed) in combina tion with a larg& cold lap opening (e.g.,
in open armam generator manway o" reactor coo 1 Ant puvap).
Und er-these: conditions, boiling can pressucit$ the reactor vaseel and push the coolane out of the core through the cpaning in the cold-leg This can l
be avoided by not isolating all the hot legs with nozzle dams, not closing all isolation valves, or by_providing an adequate hot-lag ~ vent path.
1 l
Level indication Problems:
L peauper water level' control ruquires accurate and reliable level indication.
Water level indicating systemn that can be collapsed by vacuum condfefons or
- t. ha t are easily kinked are not reliable.
Improper ' lineup of the level indicating system, inadequate installation and/or venting, ar.d l un f amil ';a ri ty wi th level indicating system.responsaw.u n lead to water level indication and cer. trol pr ob1 wn.
At 9.n o y plants, flexiole vinyl tubing used to monitor cw+ +" <essel water level is
- I
installed and suppurted by draping it over various coroponen t s.
Conumquently, the reactor vewwel water level indicatton can bw disturbed by. pas'41ng perswdnnel, the for'mation of a i r-bubbles in the liquid-filled portion, o r-icop weals in the vented portlon, r
in inany of the even ts d iscussed, the recurring problerns wfth reactor vessel water lovel Indicating systema uroded the operators' confidence fn the accuracy of the indications, In s onna cewex, operators cont (nued draining the enac t ort v ew s. e l although they suspected the indicated water level was incornwet.
In the May 1988 ela ter ford 3 event, operators continued to drain the reactor vessel with a wide disparity bwtwsmo the two water-level moni toring sys terna.
The operatoes chows to rely ori the higher level indicating system rather than thu T o w e r,, more conservative level indication.
As a result, the reactor vessel was drained until a loss of shutdown cooling occurrud.
The lower indication was later determined to be enore representative of actual reactor vessel water lavel.
Control of Draining Evolution:
Operating at reduced reactor vessel watee levels requires continuous operator attention and a timely response to residual heat removal system problems due to the reduced water inventory.
By being knowledgeable of all activities that could reduce reactor coolant inventory,-operators can quickly identi f y and cor rect conditions affecting reactor vessel water inval.
Because monitoring for unexpected drainage depends l
primarily on the level detration system, stopping any intentional draining whenever the level detection system -i s I
lost or the accuracy becomes suspect will reduce the risk of j
lowering the level below mid-loop.
Attention ta also needed to prevent conditions that could lead to the inadvertent loss of water 'from the reactor coolant system.
In the 01cblo Canyon event, a drain valve was opened without operator knowledge and was not monttored local,1y-for continuing leakage.
Therefore, when l
water level decreased, the' plant ~
the volume control tank operators could-not properly j
diagnose tha probine.
Inappropriate action was taken that i
eventually resulted in a loss of restdual heat removal I
cepability and boiling in the-core.
In the July 1986 Waterford 3 event, too drain paths were in use, but one was not incorporated in the proc 6 dure used for draining.
When securing g
i the draining evolution, only one_ drain path-(the one addressed in the procedure) was secured, and, unknown to the operators, reactor coolant level' continued to decrease until shutdown cooling was lost.
l l
Core Heatup Rates /7emperature Indications:
Heatup rates I
are a f f acted by inany faerors (e.g.,
power history, water,fnventory, time since Shutdowei, and cool ing wa ter'
??
i t ernpera t u r es and flow rates) and should ba reviewed whenever any of these factors change.
providinG expected coactor coolan': s yo turn h'en tup ra t tna to ope ra tor's can help them plan their esspon'se in the event of a loss of resfdual heat removal cooling._ Use of at least two rneans of tomasuring core
- t. *nzipe r a t u r e, one of which is indeGodent of residual heat osmoval Flow, allows monfcocing of core heetup under all cond it lons and will r e,du c e the r(ak masocinted wit h a loss of i.oro coolinej.
In the Diablo Canyon event, operators ondereu t finn ted Ihe core hurtup rate when residual heat r winove l 44 3r lost and corv twmparatune indication was not available.
OurIng t his t f ina, conw t wenps.ra tu re increased f r orn 87 degress to
)
2 ? fi dwgrews Fahrwnheie, and reactor coolant system prussure increased from a t inos phur Ic to approx itna tely 10 paig.
Ro ll ing took place in the core, and bot h s tear, and wa ter were released l
t o con taininen t.
Core rmnpena ture mon i tor ing via c,or e thermocouples might not be feasible during evolutions such as removal oe installation of thu reactor vwswel hund.
During these evolutions, maintaining core cooling with the reactor vessel water level well above the mid-loop levels can reduce the risk of losing core cooling.
All evolutions that can 'Impa c t reactor vessel watee inventory and core cool ing characteristica need to be stopped whenever a reliable core temperature indication is not evailable.
If core thermocouple twmperature monitoring will not be available for extended time perioda, a backup means to monitor core
{
camperature needs to be providad.
Restoring inventory and-Coo 11ngs-I l
Ouring e.he referenced events, delays in restoration of core cooling occurred because tho operator's lacked: the fundamental knowledge to assess the symptoms..and understand the afgnificance of the loss-of-cooling capability and also lacked the procedures ne'eded.to respond properly to'the events, w
Diagnosis and correction of residual heat removal problems l
requires' accurate information regarding both system configurations and activities affecting reactor coolant inventory.
Once the sesam generator primary manways have been untorquod or renoved, attempts to fill the pr imary sys tem.::en result in a discharge from the manway unless proper controls are in place.
Such a discharge of hot water endangers workers in or around the steem generators.
To prevent delays in restoring residual heat removal and to avoid risks of injuries or con tarn ina tion, the capabil ity to quickly evacuate workers from the area of any reactor coolant system opening af ter any
'oss of.. residual heat removal cooling or reactor vessel level is ne6ded. 1This requires'capfd communication with personnel within the containment.
In order to rneke appropriate ducis ions regard ing restoration of inventory and coollny, operaton o ed to maintain cognizance of 23
1 4
the followings o
available makeup' water sources o
al' ternate means for Injection, including gravity feed and purnped injuctIon o_
alternate cooling schemes, such 49 natural circulation of stearn to the steam generators o
oparational concerns for each cooling and rnakeup schama 3
At leaut une source of burated rnakeup water to the reactor coolant s ys t ern is needed at all times.
Gravity feed f roen the refueling water atorage tank to a ventwd pe l tua r y s ys t ern can be an uffective rumens of providing makeup water unless steam from reactor coolant system boil-off pressurizes the prirnary system above the available gravity hund.
Punips (e.g.,
charging) will be necessary to inject the makeup watwr if the available head f rorn the refueling water storage tank is less than reactor coolant system pressure.
Oscay heat races, vont flow rates, cooling by natural c i rc:ul a t ion to the steam generators, and the amount of air in the primary system will affect the cressurization of the prirnary sys tem and, thus, the feasibility of gravity feed.
Vanting and Fi111ng Insufficient water level is evident whenever a residual heat cernove l pump shows syrnptoms of cavitating-(e.g., I'luctuating current, suction pressure, or flow), even if the level detectors indicate adequate water level.
Raducing residual heat removal flow can sometimes reduce air binding-and help maintain some conidual heat-removal flow.
Once air entrainment into t he residual heat removal system occurs and the residual heat removal pumps are stopped, water level must be raised, and the cesIdual heat-removal pump suctton mugt he vented to-restore coce cooling.
Switching or starting additional residual heat removal pumps before raising water level usually will aggravate the condition by increasing the-amount of air un t ea intnen t or by causing aie entrainment<in tha=second pump.
In the San Onofre 2 event, both low pressure safety -injection pumps becaan airbound, resulting in a loss of all core cooling-because the operators started the pumps wi th the reactor vessel water level below mid-loop.
Si rni l a r l y, in the Zion 2 event, residual heat removal cooling was lost fer approximately-75 ruinu tes because the operators improperly d-lagnosed a cavitating residual heat removal pump.
Steam can collect in the auction line of residual heat removal systems designed with loop seals in the suction line.
This steam cannot be condensed by saturated liqu-Id from the hot leg exhausted with the limited capacities-of the vacuum criming or s ys t ern used by some plants, in the July 1986 Waterford 3 avant, the operators had to rap *at.nlly Jos the low pressure 4
hb l
1 J
1 I
i l
Safety injection purnps to condense the stoam bubble that had
)
f ortued in the 16w prewsure safety injection suction Ifne, r e s ul t -I ng iri a delay in the restoration of shutdown c ool -i n g.
l Concluslon:
Thw causua of the events diwcussed in this SOER Include the following:
(
o involvumwnt by rnanague w and wupwevfwoes was not e f f wc t' I vu in providfog thw shut down plant with adequate.- reliable l
core cooling evidenced by the following as f ne f f ec tive implementation of lessons learned fenm s I rn i l a r industry events Inadequate design, installation, and use of both permanent and ternporary sym tems for indicating and alarming. reactor vessel water level and for measuring core t ernper a t u r u insufficlent procedures and lack of procedure use for evolutions affecting reactor coolant inventory Inadequa te communica.t lons among plant personnel Insufficient-operator knowledge and training in the pravention and mitigation of loss of residual heat i
removal capability proceeding with evolutfons in progrese in the face of questionable information and unexpected reactor plant behavior l
o Conteal of activities was inadequate-to prevent and quickly correct unexpected coolant _. drainage paths and to remove personnel quickly from reactor 1 coolant ~ system openings-when necessary.
RECOMMENDATIONS:
(Applicable only to pressurized water g
reactors) l Management:
1.
Ensure that administrativw controls, procedures, and level indication-and alarm systems needed to safely operate with' reactor vessel water level lowered-to mid-loop-are_to place and effective.
plant ruanagement should review with stat-lon personnel the lassons learned and. potential problems associated with reduced reactor-vessel water level prior to each reduction of:ceactor-vessel water I
level _to mid-loop.
1 Operations:
?S
f
~
2.
Rovicw the procedures supporting residual beat removal system operation to ensure that procedure improvementa recommended-in SOER 85-4 and other procedure Ituprovemen t s necessary to support plant actions in rwsponse to 30ER 85-4 have been incorporated.
En s u r's that the followfng spanific i t erns are included.
a.
.sppropriate response actions for low reactor vessel wn t e,c level cundItions and symptoms of punip cavitation b.
ruethods to estwb1fwh and maintain hot-lag vent poth a
to prevent preasure buildup in the reactor vessel u.
Instructions for enwuring that temporary reactor vessel water level indicating systems are proper'ly Installed, vented, calibrated and physically walked down before being used.to monitor roantor vessel water level d.
the water level band to be raainta'Inad when at mid-loop operation to prevent flooding or loss of res idua'1 heat removal pump suction e,
actions to be taken if a leak is suspected f,
actions to be taken if core temperature i n s t rutnan t a tion is unavailable and residua 1 heat removal flow la lost 9
ma da (such as graphs) to determine anactor core heatup and bo11-off cates function of reactor as a coolant system volume and the time since shutdown, assuming worst case power history h,
instructions to restore containment closure in situations when residual heat removal is loct f.
methods-for initiating prompt evacuation of personnel from areas in and-around openings in the reactor coolant syrtem when residual heat removal or reactor vessel water level indication is lost Training:
3.
Review initial and continuing training for operations personnel to ensure that SOER-85-4 training recommendations and other tcainIng necessary to support plant actions-in response to.-SOER 85-4 have been incorporated.
Ensure t. a in ing emphasizes lessons learned from in-house and industry events involving the loss of residual heat rwuo v a l capability, including the following:
?6 m
l'-
4.
response to discrepancias in or loss o' indicated l a,ve l.,
b.
methods to detecentne decay heatup ratew c.
indications of pump cavitation and actions needed to restore core cooling flow t
d.
response to a loss-of-core cooling flow with no Md ic.a t 'on o f c o r's coolant t stupern t u r's Design 4.
I aview the design of the residual heat t'enioval sys t em and the reactor vessel 'avel tndscation and alarm system to ensure that SOER 8G-4 r ec oturnenda t i ons have been incorporated.
Ex9ariance since 1985 corroborates the need for one of the independent reactor vessel level indicators described in SOER 85-4, t'acomtnendation 6, to provide readout and low-level-alarm in the control room and have the following additional operational characteristics:
o indication scaled for mid-loop operation coerable under conditions of vacuum and o
core boiling Based on the events described-above, ensure that the following additional design features are provided:
a.
Local vents or priming pumps have the capacity to remove air and steam that can be entrained in residual heat removal suction linea.
I h.
If core thermocouples are not operable, an alternate means'or monitoring reactor coolant temperature exists that is indicated in the control room, independent of residual-boat' removal flow and operabia before reactor vessel' water level is lowered to mid-loop.
Ut111 tics and participants are requested to provide feedback on similar occurrences and solutions at their plants or on their equipment to the-information contact listed below.
LIMITED OISTRTRUTTON COPYRIGHT 1988 SY THE INSTITUTE OF NUCLEAR POWER OPERATIONS.
ALL RIGHTS RESERVED.
NOT FOR SALE-.
UNAUTHORTIEO REPROOUCTION 13 A VIOLATION OF APPLICABLE LAW.
REPROQUCTION OF NOT MORE THFN TFN COPIES BY EACH' RECIPIENT FOR 77 j
D 4
4 iTS INTERNAL USE OR USE BY ITS CONTRACTORS IN THE NORMAL COURSE OF BUSINESS IS, PERMITTED.,THIS REPORT SHOULO NOT BE OTHERWISE TRANSFERRED OR DELIVERED TO ANY THIRD PERSON, AND ITS CONTENTS SHOULO NOT BE MADE PUBLIC, WITHOUT THE PRIOR AGREEMENT OF INPO.
In f ortun t Ion Cotit ac ts :
OcganientIon and Recaturnenda t -f on 1 Adniin im t en t ion Organiratton Tv Adro in i n t ea t liin Contoet4 De pe r t ruen t (404) 953-7679 Operations
Contact:
Rec ornmenda t i on 2 Operations De pa r t roen t (404)'951-4712 Training Contactr Recomtnendation 3 Technfeal Development Oeoar tmer t (404) 953-5479 Design
Contact:
Recommendation 4 Dewtyn Engineering Department (404) 953-5417 KEY WORDS:
Core Cooling, Oecay Heat Removal, Residual Heat Removal, Reactor Vessel Level I
i l
l l
ATTACHMENT 1 MOOTFICATIONS TO ZION'S REACTOR-VESSEL WATER LEVEL INDICATTNG SYSTEM l
i To iraprove the accuracy of reactor vessel water level Indicatfans et Zion, modiffcations were madef(or are planned) l to both waree level indicating systems.
Major modifications performed' Include the following:
o.The connection points for both level indica ting sys tenis will be relocated' The'new connections will.be located at the reactor coolant system. low points.
In addition, both level indicat'fons systems are'now vented to the pressurizer.
Both systems will' remain installed and isolated from the reactor coolant system during unit operation.
The Installations have been seismically l
designed.
A' new narrow-range transtaitter has been added to the "efueling water level -indicating system for-greater accuracy siene t he rnid-loop range.
28
e To upgrade the flexible vinyl i
o entice' system was hard-piped. tubing system, the
)
(polycarbonate)
A hard plastic was uwud for a Sight glass."
All other sections of the vinyl tubing were replaced with stainless steel pfpfng cod tubing.
This s ys t ern connects to a loop drain ifne of a wteam generatur.
immediately downstream u
The electronic system for the refueling water level teenwmitter was elao upgraded.
An incore flux dwtector guido thimble tube is now used as the i
connwetfon.
The elbow connectton has been enade below thw Saal table between-the t
second conduit weld and Ihw mwel t4bla.
Most of weettons of conduit the netaalning unuawd were left in place so they may be nas11y recont. acted globe valves if necessany.
Safety-related were installed on the new connection to provida double isolation from system pressure boundary.
the-reactor coolant I
Tha following additional actions were taken to impeove the reliability of the level ind ica t ing sys terns Sensing and venting l ines for tha-level indicators o
sloped continuously were to avoid intermediate high and low points that may cause air bubbles and loop seals.
A walkdown was performed conformance with design. requirements,t,o verify design d'Imensions.
Paeticular including verifying the slope of the attention was given to sensing.and venting lines and to determining the exact level of the transmittore fer calibration purpos es.
An automatic blowdown o
trap was refueling water level installed on the moisture accumulation. indicating system to eliminate High and low-level alarms o
were-provided in the conteol room, with set points based on plant-specific data on residual heat flow rates and reactorremoval operation at various vessel water levels.
1 The refueling water o
petor to initial use andlevwl' transmitter was calibrated is recalibrated at the beginning of each refueling outage.
The transmitter output hard-piped visualwill be cross-checked regularly wfth the-level' indicator.
The calibrations take into consideration the temperature differential between the transmitter and the~ reactor. coolant temperature, water Additional details on the modifications i
vessel water level indica t ing sys t win to Zion's reacton ruede i
new provided in Reference 04
..e-.
w