IR 05000267/1987026
| ML20236G063 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/29/1987 |
| From: | Callan L, Ireland R, Jaudon J, Skow M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20236G023 | List: |
| References | |
| 50-267-87-26, NUDOCS 8711020397 | |
| Download: ML20236G063 (27) | |
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c.4 , " APPENDIX ' ju -
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REGION-IV ,, l
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AUGMENTED INSPECTION TEAM <
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.J Report No.: 50-267/87-26 ' License: -DPR-34 ' S. .Qc?
- Docket:No.: -50-267-- .
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' Licensee: >PublicServiceCapanyofColorado(PSC) ,,
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2420 W.'26th Avsn.n, Suite 15c .i ; * ( Denver, Colorado 80211 , ,
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Facility 'ame: N Fort St. Vrain Nuclear Generating. Station (FSV)
. Inspection At: FSV' Site', Weld County, C41orado ; , j. ; b .
Inspection Conducted: g0ctober 3-6, 1987 i
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ers: a /M lO 18 $) J. f. J{udon/ Chief, Project Section A TJate is
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DiMsion of Heactor Projects q; 1 ,
, koodina R.'E. IrEtend, Ch'ief, Plant Systems Section Jd2e&7 Dat6 '
Division of, Reactor Safety-i ' o
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fA ) il g . ow, React 5r Inspector D~ath J
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Accompanying- R, ae, - ,
'ji Personnel: K. L, Heitner, Project Manager, NRR D. P. Notley, NRR
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/j s , ;A C. R. Nichols, NRR' I
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R. E. Farrell,-Senior Resident Inspector, FSV < P. Micnaud, Resident Inspector, FSV s, ,!
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Approved: .h- /C f[27 L G Calian, Dirfector, Division of Reactor Daths /. Projects, Reg:4n IV ,
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f 5- r. . %lQ T/, W ll . 4N Table of Contents Page D,i
*fF 1. 0L General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 .u, se<> '- '!LI.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 Fire of October 2-3, 1987. . . . . . . . . . . . . . . . . . I 1.3' AIT Tasks. . . . . . . . . . . . . . . . . . . . . . . . . . 1 12.0 'AIT. Inspection. . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.'1 Plant Description. . . . . . . . . . . . . . . . . . . . . . 2 2.2 Sequence of Events . . . . . . . . . . . . . . . . . . . . . 3 2.3' Shutdown and Cooling of the Reactur. . . . . . . . . . . . . 6 i 0!i Review of.the Fire . . . . . . . . . . . . . . . . . . . . . . . 9 : 3' 1 Overview . . . . . . . . . . . . . . . . . . .. . . . . . . . . 9 3.2 Fire Detection. Syste . . . . . . . . , . . . . . . . . . . 9 '3.3 Licensing Background'. . . . . . . . . . . . . . . . . . . 11 .3.4 Fire Brigade Respons . . . . . . . . . . . . . . . . . . 12 3.5 ' Smoke Infiltration into the Control Room . . . . . . . . . 12 '
4.0' Review of Cdntrol Room Ventilation. . . . . . . . . . . . . . . . 12 4.1?0v9tvihw.....Yi ~
. . . . . . . . . . 12 4. 2~ Locatidh of Spaces . . . . . , . . . . . . . . . . . . . . 13 4.3: Design:of the Control' Room Habitability System . . . . . . 13 9 '4.4 Operation of the Control Room Habitability System L -During the Event . . . . . . . . . . . . . . .. . . . . . 14 i,
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s [s. O 1 Review.of Hydraulic System.(. . . . . i - . . . . '.
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k- :d[ 5.1 System Design and Operation During the Fire. . . . . . . . 14 h -a 5.2- F,il.ter Bowl Failure. . . . . . . . . . . . . . . . . . . . 16 Ms ,e c.
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,6. 0 Persons Contacted . . . . . . . . . . . . . . . . . . . . . . . 17 ya ]?0 Conclusions and Findings of Fac . . . . . . . . . . . . . . . . 17
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'?g'flott(St.;Vrain(FSV)Jlast'.refueledLinearly-1984.?-Durkl>g1startupMsting?'
3s+ hypower ascension (intJune 1984,Mreact'or scrantoccufre The r0actob ' W
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y' @ramshutdownthe. reactor;howeverC6'of;37 fin' trol?rodpairs;didnot; D iOa ~9 '
;jnsert on the. scram. 'Th'ese sixl rod pairs 3wereisuccessfully~ inserted by - .
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, d,,.s' dr!vinithem into!the core....-A'prolongedshetdownoccurred,;ddingwhich ,
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Wir control rod operating assemblies weref turbished. Operations were i' Tig' slimited. to:less than 8 perepnt during the fl1 of,1985?to dryout the. core '
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andLin the? spring-of/1986, to 35 percent y er.1 A major obrages tol resolve r
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Lequipment' qualification -questions: took. pla'j ., between June 1986 :and ' " w , ' March 1987. Operations then resumed,;at$the NRC authortied"poweElevels'
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LuVto;82; percent, ,
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' . Before^rengd 82-percent, the;lirEhseo,shutdownin order to replace'oneL j Af the p four j gas ng'cinulators. During the - .fV 2359 OMOT $cnOctober2,1937Q(partupp[fterthis ,
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The licerdee'hadL a'historyiof decreasinMerforicance, as. reflected in SALP-ryrks," u iing the early'1980'sJ This tr/yd pas been reversed 'as evidenced: 1by the to recent SALP.- :It'wi l
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' not tie 1 judged as a weak area"g inrfscentnoted,t' At fire. protection SALP report .-
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cl.2 Fire oV 0ctober 2-3, 1987-
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..! :4 A fire occurred in the turbine buildingLlate.on October 2, U 8 It was, extinguished"early on October '3,11987. . lhe high pressureihydraulictsystem played a significant.part in this fir T% system'is the likely source' '
x Lof.thecombustiblematerial',andhydrauMWoi(certainlysustainedthe'. # firc and provided the dense smoke, which 7ilpthe turbine buiIding and,, _to L a:1 esser ex@nt,;the control roo ,
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Region IV formed.'an: Accident Investigation Team-(AIT)~on October 3,'198 , The' team was ' supported by NRR,.who supplied three. members.< The AIT tasks -
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were specified in the charter letter. h ose' tasks were:
" Develop a detailed sequenced o events relating to the' fir " 2. - Verify that thel plant is in'a safe condition, with.-'a. reliable means 'of removing decay heat and with adequate and reliable instrumentation for. monitoring plant condition ,, " Determine the probable root cause of.the fire and verify that the .
licensee has taken steps-to preclude further hazards resulting:-from ' the effects of-the fir ' e
"4? Determine the'effect the fire-had on plant operation, including its effect on. plant' instrumentation and controls, equipment operability
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.(both safety and non-safety related equipment), the control room indications-and alarms and their effect on use of emergency operating . procedures by the control room operator " Determine the' facts related to the fire that are necessary for . evaluating the event's consistency with: (a) the Fort St. Vrain design basis and (b) the fire protection measures established by 10 CFR 50 Appendix R."
2. 0 AIT Insoection 2.1 Plant Description
~A brief description of the Fort St. Vrain reactor plant is provided-because of its unique design. The description is not all encompassing but is to provide assistance for readers who are not well versed in HTGR technology and terminolog f . Fort St. Vrain is a high temperature gas cooled reactor. The primary l coolant is helium. The helium flows through drilled passages in graphite moderator blocks; these blocks also contain the fue '
The hot gas coolant then flows to steam generators. There are two steam generators (one per
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i loop), but each steam generator consists of six modules. Each loop also l
'has two circulators, which act to circulate the helium gas. The !
circulators are.normally driven by steam, but water from several sources ' can be used in an emergency. The reactor, steam generators, dnd .! circulators are all contained within a prestressed concrete reactor- I vessel-(PCRV). This PCRV consists of a steel liner and thick concrete I vessel, which is kept in compression by steel tendon ! i The secondary flow path is that superheated steam goes from the steam . _i generators to either the.high pressure turbine or bypass tanks, thence to' i the. circulators as a motive force, back to the steam generators for reheating, and then to the. intermediate and low pressure turbines or to f the steam bypass system. The steam returning from the high pressure 1 turbine or- bypass. tank is called " cold reheat" steam. After this steam j has gone back through the steam generators, it is termed " hot reheat" ! steam, Temperatures and pressures on the secondary side are much greater j than those found in LWR ' The reactor building has levels 1 through 11. Level'5 is grade leve The control room and the turbine are on level 7 of the turbine building, which is adjacent to the reactor building (see Figures 1 and 2). The fire 1 was on level.6 of the turbine buildin Level 6 of the turbine building J contains the hot reheat and main steam lines. These lines are in close proximity, and there are several large hydraulic valves in the are !l
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3, 2.2 Sequence of Events 10/02/87 2350 MDT Control room operators transferred to the main steam bypass from the startup bypass for Loop 2 (Loop 1 had already been transferred) by shutting HV-2292. Operators subsequently noticed Loop 2 hydraulic system pressure was low at l approximately 2840 psig (3000 psig is normal). l l 2351 MDT The turbine equipment operator.was dispatched to check for oil leak .1 The turbine equipment operator reported to the control room
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2355 MDT that oil was flowing into the catch basin (turbine building, Level 5). The operator then proceeded to Level 6 to determine the source of the oil, which was expected to be a relief valve on a hydraulically operated valv MDT The turbine equipment operator reported a fire around HV-2292. (The fiume was later described as being 3 by 8 feet.) The equipment operator then emptied one dry chemical fire extinguisher onto t?e fir This extinguished the flames, but the equipment ope ator was forced to retreat from the area due to heavy smoke 4.nd continuing spray of hydraulic oil. The local hydraulh: isolation valves could not be reached. Fire alarm was sounded from control rco The spraying oil reflashed into a "large fire."
NOTE: The spraying oil was a separate occurrence from the oil flowing into the catch basin from a relief valv /03/87 0000 MDT The reactor equipment operator was dispatched from the control room to isolate the hydraulic supply to the Loop 2, Group 1 header, which provides oil to HV-2292. These isolation valves are on level 1 of the reactor buildin MDT Control room operators, knowing that valves for "D" helium circulator are also supplied from the Loop 2, Group 1 hydraulic header, started lowering "D" circulator speed in preparation for taking it out-of-servic Simultaneously,
"C" circulator speed was raised in order to maintain flow in Loop MDT The shift supervisor, after discussion with the fire brigade leader, requested outside assistance to fight the fir .. .
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0006 MD "C" circulator tripped. The exact cause of the trip was not determined, but it was possibly the result of increasing its speed too rapidly. Loop 2 was effectively shut down because
'!C" circulator had tripped and "D" circulator was sel turbinin MDT Loop 1 shut down due to hot., reheat activity HIG (It was later determined that this was the result of a loss of power to instruments because of the fire.) Control room ventilation automatically shifted to the minimum makeup t mode (this was also the result of the fire burning instrument cabling).
0009 MDT The control room operators inserted a manual scram because of the indicated loss of primary and secondary flo MDT ' Control room ventilation was shifted to purge mode in' attempt and to clear light smoke that was coming under the ! doors between the control room and the turbine buildin Air masks were broken out by control room operator MDT Placed "B" feed pump (electric) in service because of an indication that "A'" feed pump (steam) was' not maintaining pressure because of loss of steam following the' scra MDT Hydraulic oil was isolated to Loop 2, Group MDT -The fire was extinguished, but there was heavy smoke in , turbine buildin MDT- Operators tripped dA" feed pump to conserve' stea MOT Control room operators determined the Loop 1 shutdown was' the result of a loss of power to.the hot reheat activity
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monitor Operators pulled the instrumentation modules, reset the loop shutdown signal, and established secondary l flow on Loop 1. "A" helium circulator was placed in service I on emergency condensate, at 1300 rpm (approximately 2 percent primary flow).
0025 MDT The Platteville Fire Department arrived on sit MDT An ALERT was declare MDT '"B" helium circulator was placed in service on steam at 4000 rpm (approximately 10 percent primary flow). "A" circulator was'taken off servic Secondary flow was established on Loop MDT The plant phones were not working offsite except for the lines to Greeley and to Longmont. The shift supervisor _______ __ ____ ___--- _
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l y< notified _the assistant operations superintendent, who
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initiated response team notifications via pagers from his residenc The ENS telephone remained in service.
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. NOTE: The phone problem was later found to be the result of
- . a burned cabl MDT The Colorado Department of Health was notifie l
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0050 MDT The Weld. County Emergency Center was notifie Initial-onsite accountability; two persons were not located.
, 0055 MDT All personnel accounted fo . MOT The NRC senior resident inspector was notifie MDT The NRC Operations Center was notifie MOT The local hydraulic oil isolation valve for HV-2292 actuator , was shu (This valve was_in the fire area and had not been previously accessible.). 0135:MDT The reactor equipment operator was dispatched to return Loop 2, Group 1 hydraulic header to servic ~MDT The Loop 2, Group 1 hydraulic header was returned to , service. No leaks were observed'at the fire scen j
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0158 MDT The Technical Support Center was declared operationa { 0335 MDT Commenced reorificing to balance core flo MDT The Forward Command Post (equivalent to an E0F) was declared
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operational, q 0400 MDT Orifice valves set for equal core flo MDT Completed verification that there were two independent safe shutdown paths available for maintaining core cooling. The actual cooldown was being accomplished by normal cooldown mean l 0729 MDT The core outlet temperature was below 200 ' 0815 MDT Downgraded from ALER , . i n _ . _ _ - - - - _ _ - . _ _ _ _ - _ _ - ---
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4' ascension at 26 percent power. Main steam 50 F in both loops (Loops 1 and 2). This is sam flow is redirected from the startup - l main steam bypass flash tank. All four perating, powered by nuclear steam, and to both steam generator . , One of two rating supplying steam to the 150 psi steam f motive power for the helium circulators).
)re reaches 760 F but before it reaches ered valve on the secondary side for each main steam flow for that loop from the to the main steam bypass flash tank. This accomplished for Loop 1. When loop 2 '
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rature of 760 F, the operators accomplished o tartup '2.
. bypass flash tank to the main steam
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- operators monitored hydraulic oil system ~
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c pressure usually recovers in a short -- - Ae
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control room noted that the pressure
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k,.but the hydraulic pressure did not
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dispatched to check on HV-2292 and l' sid leak and fire. The fluid leak could .
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ice the hydraulic isolation valve was in a was then made to isolate the leak from a ic fluid pumping station in the reactor - !w this would cause loss of hydraulic
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e leaking valve (HV-2292). One of the . is hydraulic fluid header would be the
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o nt tripping of Helium circulator "0", one i o y coolant Loop .
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c operators reduced Circulator "0" speed o
- condition in which the circulator is rotates because of bearing cooling
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- or mode with th circulator's flow block protective system interprets a self "
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2.3 Shutdown and Cooling of the Reactor 2.3.1 Initial Conditions The reactor was in power ascension at 26 percent power. Main steam temperature had reached 760 F in both loops (Loops 1 and 2). This is the point at which the steam flow is_ redirected from the startup bypass flash tanks to the main steam bypass flash tank. All four I helium circulators were operating, powered by nuclear _ steam, and-
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there was feedwater flow to both steam generators. One of two auxiliary boilers was operating supplying steam to the 150 psi steam l header (a backup source of motive power for the helium circulators). 1 After main steam temperature reaches 760 F but before it reaches-800 F', a hydraulically powered valve on the secondary side for each loop operates to redirect main steam flow for that loop from the startup bypass' flash tank to the main steam bypass flash tank. This
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operation had already been accomplished for Loop 1. When loop 2 reached a main steam temperature of 760 F, the operators accomplished l the switch over from the startup bypass flash tank to the main steam bypass flash tank for Loop During this operation, the operators monitored hydraulic oil system pressur A pressure drop was expected when a large valve was operated; however, hydraulic pressure usually recovers in a short time. The operator in the control room noted that the pressure dropped from a nominal 3000 psi to about 2840 psi, when Valve HV-2292 operated to switch Loop 2 from the startup bypass flash tank to the main steam bypass flash tank, but the hydraulic pressure did not recover to 3000 psi The equipment operator was dispatched to check on HV-2292 a_nd discovered the hydraulic fluid leak and fir The fluid leak could not be isolated locally, since the hydraulic. isolation valve was in the fire area. The decision was then made to isolate the leak from a larger header at the hydraulic fluid pumping station in the reactor building. The operators knew this would cause loss of hydraulic fluid to move valves than the leaking valve (HV-2292). One of the consequences of isolating this hydraulic fluid header would be the loss of control and subsequent tripping of Helium Circulator "D", one of two circulators in primary coolant Loop In anticipation of this, the operators reduced Circulator "D'.' speed to self turbinin This is a condition in which the circulator is not supplied motive force but rotates because of bearing cooling water flow. The resulting motion is not sufficient to circulate primary coolan NOTE: In the low power reactor mode with the circulator's flow block valve open, the plant protective system interprets a self turbining circulator as an operating on _-_
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While reducing speed on Circulator "D", the operator simultaneously increased the speed of Helium Circulator "C", the other circulator in Loop 2, to keep loop flows and main steam temperatures between the loops in balance. The SRO cautioned the'R0 to expect a water turbine trip-alarm at 8800 rpm. This is a normal alarm'and normally of no consequence under these conditions, since the circulator was being , powered by its steam drive turbine and not by its. water turbine. The ! operators received the water turbine trip at 8800 rpm, but also received an unexpected, and as yet, unexplained, steam turbine trip of helium Circulator "C" at 9100 rpm. The licensee is still investigating the cause of the steam turbine tri ~ 2.3.2 Interruption of Forced Circulation (IOFC) The trip of "C" circulator with "D" circulator self turbining put Loop 2 in a cot.dition of no primary coolant flow, but the
"D" circulator self turbining told the' plant protective system that Loop 2 was still operating. (There isfa plant protective system interlock which prevents both loops from being shutdown at the same time.)
At 0008 MDT,' October 3, 1987, 9 minutes after the fire was' identified, the operators received a Loop 1 shutdown. A' loop shutdown terminates both primary and secondary. flow in a . loo _ At 0009 MDT, the reactor was,in an 10FC. With the reactor critical at 26 percent power and no forced circulation, the operators inserted a manual scram. All rods inserted'normall i The operators were, at this time, losing some control re m ! indications and'some valve controls as a result of the fire burning i cable '
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2'3.3 Restoration of Forced Circulation
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i i The operators did not immediately recognize the cause of the Loop 1 shutdown. They attempted to restore Loop 1 to service, but could not get the necessary valves to ope The operators then attempted to restore Loop 2. -During this attempted restoration, the operators started the electric driven feedwater Pump "B", and tripped the steam driven feedwater Pump "A". 1 This was because the "A" pump was not maintaining feed pressure because available steam pressure decreased following the reactor tri , l The operators could not get the Loop 2 feedwater flow control valve to : open and were unable to restore Loop 2. (It was subsequently learned ! that the flow control valve would not open because of cable damage from the fire.)
At this time, the SRO realized that the plant protective system was l interlocked to prevent both primary coolant loops from shutting down L
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E, jat..thessame time;;the;operatorstattempted to' swap loop shutdown: K1 * ' signals by manually.~ resetting'the loop' shutdown. signals and inserting ,)
~a: Loop 2' shutdown signal before~ releasing.the_' reset on Loop 1 1
- shutdown. _ . Loop 2;would not~ accept a loop shutdown ' signal. ' Thef j g operators:then realized that, Loop 2 did not have a loop shutdown- y signal,;since "0"7 circulator was still self turbini_n '
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At this point, the SRO..noted-that two of the three Loop 1> reheat
~s team activity monitorsLin the plant protective system (PPS) were i ,- dead with no lights or other indications showing. cThis would cause 1 1the PPS to shutdown' Loop .
The operators manually removed the Loop 1 activity _ monitors from'the PP This removed theyLoop 1 shutdown actuation and allowed restoration'of Loop: The.operatorsthen:restoredfeedwitterflowtotheLoop1 steam
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Egenerator and restarted. Helium Circulator "A" o.. emergency condensate'
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to the Pelton (water) driv The 10FC was terminated at 002 . The- total' duration' of - the Linte'rruption of forced circulation was . . 12 minutes. The_1icensee's_ procedure-for. calculating acceptable IOFC duration < based on initial reactor. condition's-yielded a duration of-
-18 hours.. The' design basis 10FC allows.5 hours to restore flow; /however, if ' low is_not' restored within 2 hours, venting of the primary coolant is require '2.3.4 Cooldown i 'At ; 0032, the "B" helium circulator'(the second circulator in Loop l')'
Lwas placed iniservice on steam drive and the "A" circulator was
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placed in a self turbining conditio )The operators continued cooldown observing.the decay of main steam temperature as indicative that the' plant was cooling.dow ,.l The cooldown was accomplished using normal equipment and flow path s The licensee and NRC resident' inspectors both verified that two
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m independent Appendix R safe shutdown flow paths;and trains of-
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equipment were available should they have been neede The safe ; shutdown trains were not required to cool the reactor,
, The NRC resident inspectors reviewed operator actions during the event ~and observed subsequent control room operation throughout the night as the reactor was cooled. 'At 0729 on October 3, 1987, the j reactor' outlet temperature was below 200 '-
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I 3.0. Review of the Fire
. Overview '
A fire occurred at approximately 11:59 p.m. , Friday, October 2, l'987 in the Turbine Building at Fort St. Vrain (FSV).
. The fire was the apparent l result of a filter bowl on a high pressure (3,000 psig) hydraulic system ' failing and spraying petroleum based. hydraulic fluid on hot steel valve The failed filter bowl was part of the actuating mechanism for 1
. hydraulically operated valve HV 2292. The oil.from the failed bowl-sprayed onto steel bodies'of the hot reheat safety relief valve The top ;
.of these valves cannot be insulated because of their physical and
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operational characteristics. The ignition source for the oil spray released.from the failed filter bowl was the uninsulated steel of one or more of these valves. The filter bowl is believed to have failed during * operation of the loop 2 startup bypass isolation valve, which is operated while bringing the plant on line. (The reactor was operating at approximately 26 percent.of' rated power, and the operators were preparing to start steam flow to the turbine to synchronize the generator to the grid.) Several minutes after valve HV 2292 had been operated, control room !' personnel noted that hydraulic system pressure had not returned to normal
'(3000 psig). An operator was sent from the control room to investigate ;
the cause of the low hydraulic system pressure. He discovered the fire, ' immediately reported it to the Control Room, and made an initial attack 4 with a dry. chemical, hand held extinguisher. The operator was successful ; in e' extinguishing the. fire but was not able to reach the local control valve to shut off'the oil supply. Oil continued to spray from the ; ruptured filter bowl, and the fire was quickly reignite This fire was + three to.four times larger than the original fir The operator, who wa the Fire Brigade Leader on duty, immediately notified the Control Room of_ >> this_new development and' requested the mutual aid response from the Platteville Fire Department. He then went to get full protective clothing and equipment, including self-contained breathing apparatus, and. returned with a fire hose to extinguish the new fire. In the meantime, otaer fire i brigade members had assembled and were approaching the fire from the i opposite side with another fire hos J Concurrent with these activities, another individual was oispatched to J Level 1 of the Reactor Building to close the group header control valves j on the hydraulic system to stop the flow of hydraulic fluid from the
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ruptured filter caniste The operator was able to immediately close one ! of the two valves required to isolate the syste However, the handle was ! not attached to the other valve, requiring the operator to leave the area and find a wrench to operate the valve. (See paragraph 5 for additional t details.) After the oil flow was stopped, the fire brigade extinguished the fir , 3.2. Fire Detection System Fort St. Vrain has extensive fire detection capability throughout the j
. plan Three systems are of _ particular interest as a result of this fir I l
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l An " oil mist". detection: system is installed on Level 1 of the Reactor' l Building in.the vicinity of the~high pressure pumps and accumulators that mainta.in the' working pressure in the hydraulic system. This detection- i system malfunctioned during_the day of Friday, 0ctober 2, 1987, and had
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, , .given false. alarms at about:15-second intervals. .Sometime Friday afternoon !
o - the annunciator (audible alarm) in the Control Room'was turned off.- i
~ A " linear beam'.' detection system, which utilizes. an invisible laser 1ight -{
beam, and~an ionization smoke detection system are installed in the ,
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turbine building where the fire occurre i i
- TheT11near beam detection system is relatively new and utilize " state-of-the-art" technology. The licensee has' experienced more orLless
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continuous problems'with'the installation, including inadvertent 1 interruption of the invisible beam by; plant personnel and difficulties- : with maintaining proper. alignment of the. system optical components'. I , order'to assist with identification of the obstructions to the laser beam,
-the: licensee hasLinstalled'a separate parallel visible light sourc There is also~an-ionization detection system. There were no particular ;
problems reported by; the licensee,regarding the ionization detector l There are'three separate alarm panels in the Control Room complex associated with the several fire detection. systems throughout the plan Panel.No,' 1 is' associated with-the KIDDE automatic detection / suppression j
, system located in Building 10,.and-is4 not'of interest with respect lto this . - fire. investigatio i Panel No. 2 is.a new, Gamewell panel and-is' associated with the linear * -beam detection-system installed in the Turbine Building and the Reactor !
Building, and the ionization smoke detectionfsystem located in the Turbine 1 Buildin Both Panel 1 and Panel 2 are located in an isolated area of the Control Room complex, behind a.normally closed door.at the end of a short hallwa Both panels alarm only with indicator lights-to show which system and/or ,
, zone of a system has alarmed. All audible annunciation is through I Panel 3,. located in the control room by the "SR0 desk."
Panel 3 is the original Pyralarm panel located in the Control Room proper i near the SR0's desk. .It contains some individual system and zone alarms in addition to the only annunication capability for all alarms associated ' with the plant fire protection (detection and suppression) systems. When
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a system that is monitored by either Panel 1 or Panel 2 alarms, it will j
, indicate by lighted window on Panel 3 and will annunicate at Panel ;
This arrangement is consistent with, and satisfies, the NRC fire t protection guidance and. requirement When the annunicator was silenced i I I r f __'_.L.__ _ _ _ - - . - _ _ . _
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11 li i i a, (turned off)'because of the' continuous false alarm failure of the oil mist ) detection system in the Reactor Building Level 1, audible annunciation
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capability for the various fire protection systems in the plant was los )
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3.3 Licensing Background The licensee and NRC both recognized the potential hazard from the hydraulic fluid in this are In their letter dated May 19, 1976, the licensee gave a detailed evaluation of the. combustible hydraulic fluid (Gulf Harmony No. 53) including ignition temperatures on bare metal surfaces and oil-soaked insulation over high temperature metal. The licensee also evaluated the potential benefits of changing to a synthetic
" fire proof" phosphate ester hydraulic fluid. The licensee concluded that potential problems outweighed potential benefits of switching to a phosphate ester fluid, and that hazards. associated with the continued use of Gulf Harmony No. 53 could be controlle ,
l In their May 19, 1976, letter, the licensee. committed to either install shields to prevent potential oil' spray from contacting hot or insulated surfaces, or assure that exposed metal surfaces are always below the ignition temperature of the oil spray. Such shield were installed on all hydraulically operated valves. There were no. shields protecting the hot reheat safety relief valves, which are not hydraulically operate . The only similar arrangement found by the NRC inspection team was for the main
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steam relief valves. These valves are located in the reactor buildin There did not appear to be hydraulic valves in close proximity to the In their letter dated November 13, 1978, the licensee transmitted I Section 4.0 (the fire hazards analysis) of their report titled " Fire Protection Program Review for Fort St. Vrain Nuc'. ear Generating' Station in ; Response of Branch Technical Position 9.5-1." In that Section 4.0, the j licensee described a postulated fire involving leaking hydraulic fluid i contacting insulated steam pipin The licensee c'oncluded their potential fire description with the-following:
"To ensure the early detection.and rapid suppression of such a fire, the
, following actions will be taken:
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Fix fire detectors will be added above the eight hydraulic operated high temperature steam valve " Supply oil isolation valves are provided upstream of each steam valve j hydraulic operator. An investigation will be performed to determine 1 if the supply cil isolation valves are accessible from a safe location during a fir )
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Two additional CO 2 hose reel stations will be added to the existing Turbine Building CO2 system to provide high capacity backup for the portable extinguisher Each station will be equipped with 100 feet ! of 1 inch diameter hose."
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These' modifications committed above were made, and the staff accepted the i evaluations'and descriptions contained in both of these letters and did not i seek any additional protection or modification .; 3.4 Fire Brigade Response The individual sent from the Control Room initially to investigate why pressure had not returned to normal in the hydraulic system following actuation of the hot reheat safety relief valves was a member of the FSV fire brigade. His immediate notification of the control room followed by i
. the prompt response to the fire reflected an apparently good training !
progra The initial extinguishment of the fire without first eliminating j the oil flow from the ruptured filter bowl indicated a gap in the. training I program and probably contributed to increased fire damage because the larger fire that resulted from reignition of the hydraulic fluid spra .l
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3.5 Smoke Infiltration into the Control Room Large quantities of smok'e were produced by this fire. There was sufficient smoke infiltration into the Control Room to r.equire use of the piped in breathing air and masks provided. (The Control Room HVAC system ] and how its operation contributed to the smoke' infiltration is covered in - paragraph 4 of this report.) Smoke migration throughout a plant during a fire emergency is covered only in general' ways in-the'NRC fire protection guidance documents. Section D.4, Ventilation, of Appendix A to BTP ASB 9.5.1 states that,' "The products of combustion that need to be j removed from a specific fire area should be evaluated to determine how . they will be controlle Smoke and corrosive gases should generally be automatically discharged directly outside to a safe location." 1 (Section D.4.(a)). 4.0 Review of Control Room Ventilation j
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4.1 Overview
During the October 2-3 fire in the turbine building at the Fort St. Vrain i Station, smoke from the fire quickly entered the control room atmospher The smoke in the control room atmosphere persisted even after manual transfer of the control room HVAC to the'" smoke purge" mode and remained for another approximately 5 to 20 minutes until purged by inflow of clean ' air after the door to Building 10 was propped ope Personnel in the
' control room used the Breathable Air System to enable them to continue working in the heavy smoke atmosphere of the control roo ,
The investigation of the control room habitability aspects of the fire event focused on determining the design and operating characteristics of q the control room habitability system and how they related to habitability conditions during the event. Based on the cause and effect relationships ] so established, the investigation sougV to establish the basic or root
' ~ cause underlying the response of the control room habitability system to the fire event, j
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(control room ventilation equipment) in the Turbine Building. The north (
end of.the area above the turbine building floor is-immediately adjacent to the. south wall of the control room (see Figures 1 and 2). There are four ; access doors to the control room. Two allow direct passage from the control room or the reactor engineer's office to the area of,the turbine building floo One allows direct passage.from the control room to Building 10. Another allows' passage from the control room to a walk-through structure from which passage is allowed through doors to the , turbine building operating floor or to the Building 10-training room. The
. normal fresh air intake to the control room complex is located on the roof of the Turbine Building near the' northeast corner. The emergency fresh air intake to the control' room complex is located in the-Turbine Building.at the north end of the area above the turbine building floor at an elevation above the elevation of the control room comple .3 ~ Design of the Control Room Habitability System ;
The purpose of the control' room HVAC system is to maintain a thermal environment and air quality sufficient'for persconel comfort and safety under all plant operating conditions. The system serves three separate areas: control room, reactor engineer's office, and auxiliary electrical equipment roo The system is not connected to the turbine building HVAC syste During normal operation, a limited quantity of fresh air is directed into l the system, filtered, and mixed with the recirculated. air (21,160 cfm), which is heated and/or cooled and humidified and/or dehumidified in'the air' handling units. Temperature controllers for the three areas regulate
, the air temperatures in these areas by positioning the mixing dampers *
associated with the air handling units or the mixing-dampers which control , the quantities of outside and return ai Normally, these latter dampers l are positioned so that only the minimum amount of outside air is mixed i with the return air,' A differential pressure controller modulates the
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control room exhaust air. dampers to maintain a slightly positive pressure i in the Auxiliary Electrical Equipment Room with respect to the Turbine ! Building (see Figure 3). J
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During abnormal operation following detection of a radiological event, the ! isolation dampers are aligned so that the control room HVAC system j operation is in the normal mode except that make-up air (480 cfm) is drawn l from the Turtiine Building and passes through the make-up ventilation l i j i
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filter befo/e entering the air handling uni A differential pressure controller operates a damper upstream of the filter to maintain positive control room pressure. Positive pressure in the control room minimizes the ' chance of unfiltered air inleakage to the control room (see Figure 4).
In the event that a fire is detected in the control room or Auxiliary Electrical Equipment Room, the isolation dampers are closed and the j control room HVAC fans are tripped. After the fire is extinguished, the control room atmosphere may be exhausted of contaminants by operating the system in the purge mode. This opens the isolation dampers, starts the-fans, and supplies 100 percent outside air and exhausts to the outdoors -
(see Figure 5).
The Breathable Air System is designed to provide a total of 20 scfm of I filtered outside air to support the use of six masks in the control roo The filters remove chlorine and other noxious gases. Also six Scott Air Pacs are located in or immediately outside the control roo ,, 4.4 Operation of the Control Room Habitability System During the Event The control ' room HVAC sy: tem was operating in the normal mode of operation immediately prior to the event (Figure 3). Throughout the event, personnel used the doors on numerous occasions to pass back and forth from the control room and the area above the turbine building floo About 10 minutes after the fire was detected, the stack radiation monitor failed because of a loss of power, when the power cable was burned. This caused a transfer of the control room HVAC system from the normal mode to the abnormal radiological emergency mode (Figure 4). Two minutes later, the control room HVAC system was manually transferred to the smoke purge mode to clear the light smoke condition in the control room (Figure 5).
However, the control room atmosphere filled with more smoke. The personnel in the control room began to use the masks of the Breathable Air System. Since only three masks were available, the six men had to share the use of the masks. In the next approximately 5 to 20 minutes, the door to Building 10 was propped open, and the control room was quickly purged of the heavy smoke by the inflow of clean air from Building 1 After the event, heavy deposits of smoke particles were found on the control room floor at the bottom of the door leading to the area above the turbine building floor and in the turbine buildin No smoke deposits were found near the HVAC outlet ducts into the control room working area. A damper in one of the three ducts distributing air to the control room working area was found, after the event, to be stuck in a position one quarter ope .0 Review of the Hydraulic System 5.1 System Design and Operation During the Fire i The Fort St. Vrain reactor utilizes a hydraulic (oil) system to control 30 fast acting control and block valves in the secondary coolant (steam) .-
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system.1 The' hydraulic system has a nominal operating pressure of i 3000 psi.- A separate hydraulic system exists for each of the-two c secondary. coolant loop Since rapid valve actuation can cause individual valves to demand fluid at high flow rates, the system uses both pumps and accumulators-to supply fluid to the valve actuator However, the maximum I amount of hydraulic oil available to be rapidly discharged is about 7 gallons'per valve or'13 gallons per accumulato '
The hydraulic oil used in the system in Gulf Ha'rmony 68 , with a flash
' point of 450 F. The oil was predicted not to ignite-when sprayed on hot ;
surfaces less than 820 F and to always ignite when sprayed onto surfaces i at or above 870 F. At the time of the event, main steam temperature was between 760 F and 800 Hot reheat steam temperature was, however, l approximately 850 F to 860 F. The licensee did not verify the properties l of the oil prior to'the event but was planning to obtain a laboratory i verification as 'a result of the fir .To assure reliable system operat. ion, a filter assembly is used at each valve on the intake to the control valve. (These are normally in the supply line to the solenoid operated-pilot control valve.) The. filter assembly consists.of an aluminum head and canister bowl.which contains the filter element.' The bowl is fabricated as an aluminum cylinder, 215 inches in diameter and approximately 8 inches long. The design pressure for this' ; system is 3000 psi, and the proof pressure is 7500 ps ,j The hydraulic oil leak that apparently initiated this fire was in the i filter bowl of valve HV-2292. The bowl had a longitudinal failure somewhat similar to a fish-mouth break in tubes or' pipes .The bowl was i distorted (swollen near the point of rupture). . The licensee committed to conduct a materials or metallurgical analysis of'this filter bow Oil spraying from the failed filter bowl hit hot reheat steam piping and the uninsulated (exposed) portions of the hot reheat steam relief valve .l At the time of the fire, the hot reheat steam temperature was ' approximately 860 There were no other obvious ignition sources l identifie It was concluded that the surface temperatures of the hot reheat relief ' valves caused autoignition of the oil spray from the failed filter bow As previously discussed, the failure of the filter bowl did cause a j condition that the operators recognized as " abnormal." This prompted them to examine the status of the valves that had just shifted position, which l led to the discovery of the leak and fire, j
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1 This system is described in FSAR Section 9.1 Some docketed correspondence concerning the oil referred to it as Gulf l: Harmony 53. It appeared that the name of the oil changed over the years, l but the essential characteristics did not vary significantl { i __ _ _ !
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Once the operators discovered this situation, they attempted to isolate , the hydraulic fluid (high pressure) supply to the affected valve to limit : the source of flammable material. However, the isolation valve for this hydraulic line was too close to the fir Consequently, the operators had' ! to isolate the affected valve group at the hydraulic system in the reactor ; buildin There:are no controls for performing'such an isolation ' available in the control room. In shutting the two valves to isolate hydraulic oil to the failed filter, it was found that one of the valves ; did not have-its handwheel properly installed. This was a condition " identified _in May 1987, and it had not been corrected although it was awaiting engineering action. The operator used a wrench to shut this valv After the event, the NRC inspectors reviewed the licensee's Technical Specifications (TS) and procedures concerning the hydraulic system. The focus of.the existing TSs was to. insure the system's ability to actuate the secondary coolant system valves. No surveillance requirements address leakage of hydraulic oil or its potential as a fire hazar The hydraulic system has internal differential pressure alarms which could indicate gross failures of system integrit However, the operators have-to rely on visual surveillance during normal plant operations to detect small, but potentially hazardous leaks. The only other system directed towards the leakage problem is an oil mist detector located in the reactor building near the main system pumps and accumulators.
! The licensee does not have'in place a systematic program of inservice inspection or testing to examine this system for long term deterioration or problem I 5.2 Filter Bowl Failure The NRC inspection team noted that many of the filter bowl canisters showed severe scoring on the upper exposed surface These marks were apparently caused by the use of a pipe wrench to remove the filter bowl for changing the filter elemen Licensee personnel indicated that the filter bowls are normally installed hand tight. When the bowls are isolated for maintenance, there is high pressure (3000 psig) trapped in the isolated system. There was no drain plug or other means provided to relieve this trapped pressur Maintenance personnel had therefore used the mechanical advantage of a large pipe wrench or equivalent device to remove the filter bowl The failed filter bowl had the deep grooves or score marks on it. These were at the to No score marks were found in the area of distortion and failure (which was more than halfway down the canister).
The NRC inspectors noted that the manufacturer specifications predicted filter bowl failure at approximately 15,000 psi The manufacturer
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i specifications also stated that in severe impulse service, the filter bowl i should not'be subjected to.a baseline presr,ure above 1500 psig. Since the j
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hydrau.lic system at FSV operates large hydraulic valves (e.g., 20 inches) j
.at fastLspeeds'(e.g., full stroke times of.1-2 seconds), the NRC j ' inspectors concluded that.there might be instantaneous shock waves :
reflected in the. hydraulic system which approached or exceeded the design characteristics of the filter bowls. The licensee c'id not have 1 calculations available and had not taken test data to establish the actual l instantaneous system pressure .0 Persons Contacted
*F. J.' Borst, Support Services Manager f' * L. Brey, Manager, Nuclear Licensing and Fuels *R.-L. Craun, Site' Engineering Manager M. Dennison, Assistant Operations Superintendent T. Dice, Shift' Supervisor K. Eineg, Reactor Operator D. Evans, Operations Superintendent *M. J. Ferris, QA Operations Manager * .
H. Fuller, Station Manager
*J. M.'Gramling, Supervisor, Nuclear Licensing J. Hak, Shift Supervisor P. Harmon, Technical Advisor * H. Holmes, Nuclear Licensing Manager S. Koleski, Senior Reactor Operator ; *W. Matheney, Member, NFSC i S. Murphy, Equipment Operator (Fire Brigade -Leader) * Niehoff, Nuclear Design Manager *F. J.'Novachek, Technical and Administrative'e Services Manager *T. Prenger, QA Services Manager *G. D. Schmaly,-Fire' Protection Engineer l *P. F. Tomlinson, Manager, Quality Assurance j *D.,Warembourg, Manager, Nuclear Engineering T. Weller, Senior. Reactor Operator * O. Williams, Vice President, Nuclear Operations The NRC inspectors also interviewed other PSC personnel including individuals from operations, engineering, maintenance, and health physic * Indicates presence at exit interview conducted on October 6, 198 .0 Conclusions and Findings of Fact t- The NRC inspectors reached the following conclusions and findings of fact:
A fire occurred in the Fort St. Vrain turbine building on October 2"3, 198 .
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< - , i - ' .The fire was extinguished by the site fire brigade, primarily using )
3, water in the form of fog and after the source of combustible oil was '
secured, l- hydraulic all filter bowl failed on over pressuro. This allowed ! hydraulic fluid to spray onto hot components. Whether this failure- ! was the primary cause.of..the fire or the result of some other, undiscovered failure was not determined absolutel ! The aluminum filter bowls may be'underdesigned for the service applicatio :The. maintenance practice of usinp a pipe wrench or equivalent to unthread filter bowls is not acceptable. It probably did not lead j
:the failure on October 2-3, 1987, but it could have caused a filter i bowl failure eventuall l All scored filter bowls should be either replaced or evaluated by engineering for continued us *
l The local hydraulic isolation valve was not located in a position that allowed it to be used to isolate the lea I The fire did not adversely affect the licensee's ability to shutdown and' cool the reactor. The licensee chose to conduct a'cooldown using forced circulation from a single circulator operated on steam from an auxiliary boile Heat was rejected through a steam generator fed by the electrically driven feed pum This was a faster means of ' cooldown that the Appendix R shutdown equipmen '
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Both trains of Appendix R shutdown equipment were operabl In addition to the hot reheat safety valves, which were the apparent hot surfaces which caused the oil to ignite, there.are at least two other areas within the plant which have exposed hot surfaces. These are the main steam safety relief valves for Loops 1 and 2, located in
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the reactor buildin The plant was safely shutdown and cooled. There were sufficient instrumentation and controls to accomplish the cooldow This fire occurred in an open area of the turbine building with no horizontal fire barrier Floor gratings and open areas including cable trays were below the fire,.and there was a ceiling and cable . chase above. The approximate horizontal spread of the fire area was 10-15 feet. Horizontal spread was initially to the areas where the 1 hydraulic oil was available for combustion. Thus, the fire was within the separation requirements of Section III.G of Appendix The fire could not have directly affected properly separated safe ; , shutdown systems, had the same size of fire occurred elsewhere in the I L plan j j
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The plant remained within the design basi " The operators reacted well to the event and did not lose control of y the plan l The control 1 room airline breathing system is capable of supporting
, five or six masks. There were only three masks available in the control room during the fire, but there were six people in the contro roo The coatrol room ventilation system was not designed to maintain a I smoke free atmosphere during a turbine building fire-of the type which occurre . ~ Opera. tor error in switching to the purge mode of ventilation ' contributed to the control room smok ,
Opening and closing of the control room doors for passage contributed to loss.of the positive pressure in the control roo l There was. deposition of combustion products throughout the turbine : building. The potential corrosive effects of this should be i evaluate ' Problems with the fire detection system and the turning off of the-audible annunciation for'the fire detection. alarms may have caused a short delay in discovery the fir The~ fast stroke times of hydraulic valves often cause the relief ,
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. valves for individual operators to open. This is indicative of high, J instantaneous system pressure '
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i., There are no ISI' requirements for the 3000 psig hydraulic syste The hot reheat safety relief valves and associated piping were rapidly cooled by the water used for fire suppressio ;
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