ML19345B947
| ML19345B947 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, Davis Besse |
| Issue date: | 02/20/1980 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML19345B940 | List: |
| References | |
| RTR-WASH-1400 ACRS-R-0862A, ACRS-R-862A, NUDOCS 8012020811 | |
| Download: ML19345B947 (16) | |
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Nuclear Regulatory Consnission Staf f 4
j Report, " Evaluation of Davis-Besse and Rancho Seco Feedwater Transients on 9/24/77 and 3/20/78 Using WASH-1400 Data" j
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l INTRODUCTION
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In this report we have eval- ' ad the foss of Main Feedwater transients which occurred at Davis-Besse 1 on 9/24/77 and at Rancho Seco and compared them with
'c! accident at Three Mile Island-2 on 3/29/79. A The behavior summary is provided of the Davis-Besse and Rancho Seco events.
An event tree for Loss of Main of important safety systems is compared.
Feedwater transients is provided, and each transient sequence is identified in the context of th! event tree, WASH-1400 date.
First, WASH-1400 was performed for the Certain caveats should be made.
We have not done the Westinghouse-designed Surry plant, not a B&W reactor.
Such ar kind of major in-depth analysis here that was done for WASH-1400.
Second, it should be analysis would require considerable effort and funds.
recognized that there are sig..ificant uncertainties in the WASH-1400 data.
Third, the evaluation refers to pre-TMI sy: tem behavior and transients.
DISCUSSION OF DAVIS-BESSE TRANSIENT B.
1.
Event Summary - Davis-Besse a series of events occurred at the Davis-Besse On September 24, 1977 Unit 1 which resulted in depressurization of the primary system frem a normal operating pressure of 2150 psi to 900 psi in approximately eight minutes, and the release of approximately 11,000 gallons of water in the form of steam within the containment through the pressurizer quench tank rupture disc.
On the afternoon of Saturday, September 24, 1977 the main turbine was i shut down to repair a leak in a pressure sensing connection on a steam
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The line from the turbine governing valves to the turbine inlet.
reactor was being held critical at approximately 9% thermal power.
At 2134 hours0.0247 days <br />0.593 hours <br />0.00353 weeks <br />8.11987e-4 months <br />, a spurious half trip occurred in the Steam Feedwater This caused the startup feedwater Rupture Control System (SFRCS).
valve on the No. 2 steam generator.(which is the normal feed path at Closure of this valve resulted in a low this power level) to close.
No. 2 steam generator level, which then resulted in a normal full trip SFRCS of the SFRCS for this conditior, and initiation of the SFRCS.
initiation closes both main steam isolation valves and initiates fe water flow to both steam generators from their individual steam-driven auxiliary feedpumps.
The half trip and resulting full trip of the SFRCS caused a reduction in heat removal from the primary system and a corresponding temperature /
The pressure rise in the primary pressure rise in the primary system.
This valve system caused the pressurizer power relief valve to lift.
then rapidly oscillated closed-to-open approximately nine times and The chattering of the relief valve remained in the full open position.
was caused by the physical absence of a relay in the valve control 1he relay normally provides for a deadband between logic circuitry.
"open" and "close" setpoints. An empty relay socket was found in the logic cabinet after the event.
The temperature rise in the primary system caused an increase in the pressurizer level, and the operator manually tripped the reactor on I
high pressurizer level approximately two minutes after the half trip t
on the SFRCS occurred.
The pressurizer power relief valve, in the full open position, rapidly reduced the primary system pressure, and a Safety Features Actuation i
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,.6 System (SFAS) trip occurred at the 1600 psi s:tpoint of the prirary The power relief valve discharge goes to the pressu.rizer system.
quench tank, which became overloaded and overpressurized 41/2 minutes after reactor trip the rupture disc in this tank relieved Approximately due to overpressure, venting the steam into the containment.
20 minutes after reactor trip, the operators diagnosed the reason for primary system depressurization as being the power relief valve, a the control room closed the motorized block valve ahead of the power reli.ef valve, terminating the loss of primary coolant into the contain Subsequent operator action using makeup pumps and high pressure pumps stabilized the primary system pressure and pressurizer lev controlled shutdown to cold shutdown conditions followed.
The major physical damage from the incident was to the reflective me
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insulation on the lower part of the No. 2 steam generator, which rece A ventilating the jet of steam coming from the pressurizer quench tank.
duct in the area of the quench tank was dimpled and required straighte Twenty.hree panels of reflective metal insulation required replacemen 25, 1977 Entry into the containment was made at 0550 Sunday, September for cleanup operations.
Another event occurred in the course of this incident that did contribute materially to the above events, but did result in the No. 2 This was the failure of the No. 2 ' auxiliary steam generator going dry.
This feedpump to come up to full speed (3600 rpm) following the SFRCS trip feedpump came up to approximately 2600 rpm and stayed at this lev with no flow to the steam generator until approximately 12 minutes after reactor trip, when the operators placed its control in manual anc
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brought it up to full speed (comencing feedwater flow to.the steam generator).
ystems Behavior - Davis-Besse 2.
An important fact to bear in mind while discussing the Cavis-Besse transient of 9/24/77 is that only one full-power day of operation had This means been accumulated at the time of the event (see Table 1).
that considerably less decay heat was being generated in the core than In addition, the Davis-Besse reactor was only was'the case at TMI-2.
at 9% power when the main feedwater was lost. A high pressure reactor trip did not occur (it did at TMI in 9 seconds), confirming the slower, milder nature of the Davis-Besse transient.
Although the pressurizer
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Operator reaction to the transient was effective.
level increased off-scale in the first ten minutes, the operators apparently realized the pressurizer level increase was misleading and caused by steam However, the operators did turn off the formation in the primary system.
HPI pumps (just as at TMI) after only three minutes of operation.
The The pressurizer relief valve stuck open early in the transient.
operators diagnosed this problem and closed the block valve after 21 At TMI a similar problem took 138 minutes minutes into the transient.
The ability to diagnose and take remedial action in 21 minutes to diagnose.
helped to tenninate the Davis-Besse transient with a minimum of damage.
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Event Tree Evaluation - Davis-Bessi The events at Davis-Besse on 9/24/77 can be depicted in an event tree This (Figure 1). The Davis-Besse t' ansient is f2 on the event tree.
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4 The event may be compared with sequence #3 which is the TMI-2 sequence.
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tree is for a category of transients which begin with a loss of all main In the case of Davis-Besse, this was apparently initiated feedwater (TM).
by a faulty input buffer in the logic control of the Steam Feedwater Rupture Control System.
WASH-1400 estimated three of these feedwater transients to occur per In the 12 months prior to the THI-2 accident, the year at each reactor.
aver' age number of feedwater transients at B&W reactors was three per It should be noted year (see Table 2), confirming the WASH-1400 value.
that a larger number of feedwater transients occur in the first few Perhaps 2 to 3 years of operation, and a smaller number after that.
Plants which times this number might be appropriate for early operation.
have operated longer than a few years may average 1 to 2 feedwater transie
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per year.
Within about ten seconds after the main feedwater system had tripped, increasing reactor pressure caused the pressurizer relief valve to open.
The WASH-1400 This valve then failed to close, causing a small LOCA.
failure rate estimated for this failure mode was 1x10-2 per demand with a factor 10 uncertainty up and down. More recent data in light of the TMI-2 accident indicate three relief vale failures in this mode i about 150 demands, or a failure rate (to reclose) of $2x10-2 per deman again confirming the WASH-1400 failure rate.
At the same time that the relief valve was apening in the primary system, the auxiliary feedwater system was being aligned to the steam generators A
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6-and auxiliary feedwater flow had comenced successful'y shor.tly thereaft
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About 30 seconds later, the operator tripped the reactor manually because of rising pressurizer level.
Reactor pressure did not reach the setpoint of the pressurizer safety The ECCS system automatically valves and they were not called on to open.
actuatedonlowpressure(1600 psi)intheHighPressureInjection(HPI) mode about 1 1/2 minutes after the pressurizer relief valve stuck open.
After the HPI system operated successfully for about three minutes, the Because of the nature of the transient, operator manually terminated HPI.
The probability of this was regarded as successful operation of ECCS.
this category of transient occurring in a B&W reactor, as predicted using WASH-1400 failure data, is estimated as follows:
1x10-2 3x10-2 per reactor year
=
3 x
Loss of Main Relief Valve Feedwater/yr.
Fails to Close
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DISCUSSION OF RANCHO SECO TRANSIENT C.
1.
Event Surnary - Rancho Seco an excessive cooldowa transient was experienced while On March 20, 1978 Non-nuclear instruments were operating at 70% power (IE Report 50-132).
lost including steam generator and pressurizer levels and all RCS tempera-
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Loss of RCS hot leg terperature input to the ICS caused termination tures.
Reduced heat removal in the steam generators caused of feedwater flow.
The reactor tripped on high RCS temperature and pressure to increase.
The secondary sides of both RCS pressure followed by a turbine trip.
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steam generators er ptied due to operation of condenser bypas,s valves Although normal control atmospheric dump valves and auxiliary steam loads.
room indications were lost, the computer typewriter will print alams In addition, selected plant parameters can %
setpointo are reached.
With the aid of computer in e ation, monitored on the ICS computer printout.
pressurizer level was maintained by manual operation of a high-pressu "A" steam generator level control initiated emergency injection pump.
feedwater injection (level control was actually lost at time zero, but th channel drifted slowly downward while "B" channel drifted slowly u The turbine-driven auxiliary feedwater pump had started on loss of fee flow.
RCS cooldown started as a result of emergency feedwater flow to "A" ste generator and possibly main feedwater pump flow (manually ope Decreasing RCS pressure (1600 psig) actuated HPI pumps and the motor-Full auxiliary feedwater was initiated driven auxiliary feedwater pump.
The RCS reached a minimum of 1475 psig and to both steam generators.
was then increased and maintained at 2000 psig by manual control of an HPI pump.
Restoration of the non-nuclear instrumentation restore Operating personnel secured the auxiliary feedwater pumps and controls.
and started RCS pressure reduction using the pressurizer spray.
2.
Key Systems Behavior - Rancho Seco The incident at Rancho Seco on March 20, 1978 involved a loss of main feedwater due to operator-induced failure in the ICS non nuclear e
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The incident was aggravated by the fact.that (1) the instrumentation.
f plant ICS reacted to erroneous instrument readings causing delays in initiating AFW injection and subsequently allowing excessive AFW injec and(2)theoperatorshadaverylimitednumberofinstrumentreadings which they could trust to manually bring the plant to an orderly shutdown.
Since the reactor was at 70% power and had logged considerable operatin time (31/2 years of commercial operation), the decay heat to be remove was significant, similar to TMi-2.
Auxiliary feedwater was not available for seven minutes after MFW trip.
However, this delay was not as serious as at TMI-2 because there was no small LOCA in progress; i.e., a pressurizer safety valve had opened and closed properly.
The transient was eventually brought under control by the operators' diagnosis of which electrical i.ircuit breakers had opened, and then closing them.
3.
Event Tree Evaluation - Rancho Seco The Reactor Safety Study (RSS) stated that on the average a plant can expect about three main feedwater losses of a few minutes duration per This value was obtained from the operating experience available year.
The nature of the three main at the time the RSS was in progress.
Therefore.
feedwater losses per year was not discussed in great detail.
the breakdown of the various causes of feedwater transients the Rancho Seco incident) in quantitative terms is not provided in the
~RSS.
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The NRC has investigated feedwater transients at B&W plants,an At least five of the main feedwater losses
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this information in NUREG-0550.
attributable to ICS-related failures or malfunctions were There were many Among these is the Rancho Seco incident.
I that document.
other main feedwater Esses which licensees felt were not significant It is not known how many of these were ICS enough to be reportable.
The average failure or non-nuclear instrumentation failure related.
rate of main feedwater for B&W plants subsequent to RSS was reco at three per year.
The RSS identified several potential transient-initiating events which Among those identified were are associated with the loss of feedwater.
the loss of main feedwater pumr and malfunction of control, loss of condensate pumps, loss of A.C. power to the feedwater system, and o The probability of occurrence of any one specific initiating event ma However, when assembled into appropriate categories, the net small.
In this probability of a given type of transient may be considerable.
regard, the probability of the event at Rancho Seco is a small par the larger probability that the main feedwater system will be lost.
This transient may be classified as belonging to sequence il on the ev However, this ICS/NNI initiated trans'ent could tree shown in Figure 1.
That is, the loss of NNI which resulted have been more severe than it was.
in erroneous instrument readings delayed the automatic injection o perhaps even more significant, operator information on the stat The erroneous instru-plant was severely limited throughout the transient.
ment readings eventually " drifted" to the point of AFW injection some seven minutes into the transient even th'ough the steam generator was
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O It appears that the apparently dried out by the end of the first minute.
capability existed at all times for manual action to initiate AFW injec If erroneous instrument readings or manual actions had never initiated AFW injet tion, this event would have followed the path of sequence 10 in Figure 1.
Another sequence of significance for this initiating event is sequence f3 If a pressurizer relief valve had become stuck open, this event could h been use than the TMI-2 sequence, depending on operator actions, becaus However, the of the additional problem of a lack of instrument readings.
specific initiating event ICS/NNI failure or malfunction, may be some Using WASH-1400 less likely than main feedwater losses due to other causes.
data, the overall sequence #1 would have a probability of occurrence of three times per year per plant; the specific (and potentially more seve case where the loss of NNI is the cause is expected to be a much sma of this category.
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COMPARISON OF THRFE B&W REACTOR lilCIDEilT EVEllT SEQUEllCES TMI-2 DAVIS BESSE RAllCHO SECO
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(3/20/73)
_(L/29/79)
._(9/24/77) 70%
REACTOR POWER 97%
9%
REACTOR HISTORY IN COMMERCIAL OPERATION
~1 FULL POWER DAY OF IN COMMERCIAL OPERA-THREE MONTils.
OPERAT10il.
TION 3 1/2 YEARS.
TURBINE TRIPPED IllSEDIATELY.
DOWN ALREADY.
TRIPPED.AFTER 5".
REACTOR TRIP AUTOMATIC AFTER 8" 0:!
MANUAL (1 MIN. 47")
AUTOMATIC AFTER 5" ON HI REACTOR PRESSURE BECAUSE OF RISING HI REACTOR PRESSURE.
(2355 PSI).
PRESSURIZER LEVEL.
MFW BOTH PUMPS TRIP IMME-1 PUMP TRIP IMMEDIATELY REDUCEDTOZEROFLOW DIATELY.
1 PUMP TRIP 58" LATER.
. BY FAULTY ICS SIGilAL (SOME MFW INITIATION BY OPERATOR PROBABLE AFTER 7 MIN.).
D' TM 1 (CONT 7 m
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TMI-2 DAVIS BESSE RA!!CHO SECO (3/29/79)
(9/24/77)_
(3/20/78)-
1 PUMP /SG WORKING WITHIH NO AFW FOR 7 f11tl.
46".
1 PUMP " UNAVAILABLE" (TURBINE DEGRADED).
AVAIL-ABLE MANUALLY AFTER 12 f1IN.
PRESSURIZER OPENED AFTER 3" AND OPENED AFTER 1 MIN. 6",
GAGGED CLOSED.
RELIEF VALVE STUCK OPEN.
BLOCK CYCLED RAPIDLY 9 TIMES SRV OPENED tJ1D IN 23" AND STUCK OPEN CLOSED PROPERLY VALVE CLOSED AFTER 135 (11N.
(STEM GALLING).
BLOCK VALVE CLOSED IN 20 MIN.
< PRESSURIZER SEVERELY MISLEADING LEVEL INCREASED OFF NO LEVEL PROBLEM.
LEVEL INDICATION.
SCALE.
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TABLE l OIT. )
TMI-2 DAVIS BESSE RANCHO SECO
__9/211/77)_
(3/20/73)-
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(3/29/79)
ECCS HPI AUT0 STARTED (1600 llPI AUT0 STARTED (1600 HPI MANUAL AND PSI) AT 2'02".
1 PUMP PSI) AT 2 Mill. 57" AND INTERMITTENT DURING TRIPPED AFTER RUNN!!!G PERMITTED TO RUN FOR FIRST 13 i1IN.
THE!!
2 Mill. 36".
OTl!ER 3 MIN. 5".
HANUAL AUT0 START (1600 PSI)
PUMP THROTTLED TO SHUTDOWil BECAUSE
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PRESSURIZER LEVEL NORf1AL.
It!STRUMEllTS
- MOST U.K.
0.K.
ONLY PRESSURIZER LEVEL AND RCS PRES-SURE TRUSTED BY
~ OPERATORS DURiilG FIRST 75 f11N.
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m FABLE 2
-l WASH-ll100 FAILURE RATES FAILURE RATE 1.
flAIN FEEDWATER (TM) 3/YR 2.
REACTOR TRIP (K) 3.6x10-5
/D 3.
AUXILIARY FEEDWATER (L) 3,7x10-5/D' 4.
PRESSURIZER RELIEF VALVE OPENS (P )
1x10-2/D 1
5.
SAFETY VALVES OPEil (P )
3x10-5/D 2
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PRESSURIZER RELIEF VALVE CLOSES (0 )
1x10-2/D 1
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SAFETY VALVES CLOSE (0 )
1x10-2/D 2
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ECCS - HI PRESSURE INJECTI0tl (C) 3.7x10-3 1
9.
ECCS DEGRADED OPERATI0li (C )
>3.7x:.;
' ANALYSIS UNIQUE TO SUP.P,Y G
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4 Memorandum From F. Rowrome to R. Fraley, i
"ACRS Query on Material Relevant to Udall 4
Letter: Davis-Besse and Rancho Seco Transients,"
February 12, 1980 4
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6-Attachment C
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