ML20050C173
| ML20050C173 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/10/1982 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML20050C153 | List: |
| References | |
| TASK-PII, TASK-SE SECY-82-058, SECY-82-58, NUDOCS 8204080238 | |
| Download: ML20050C173 (2) | |
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ENCLOSURE 1 TO PAGE 8 s.
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SECY-82-58 February 10. 1982
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POLICY ISSUE (Information)
The Commissioners For :
William J. Dircks Executive Director for Operations From:
MEDIA ATTENTION T5 A PRELIMINARY EVALUATION OF
Subject:
ACTIONS DURING THE GINNA EVENT To inform the Commission'of the pr'eliminary evaluation.
Purpose:
On January 28, 1982, three days after the steam generator -
tube rupture event at the Ginna reactor, the Reactor Systems Discussion:
Branch in the Office of Nuclear Reactor Regulation completed The purpose of the a preliminary evaluatiott of the event. evaluation was f
response at Ginna to the recently proposed Westinghouse emergency procedure guidelines for steam generator tube f
A copy of the resulting staff memorandum rupture events.Some of the results of the preliminary is attached.
evaluation were briefly noted by NRit (Roger Mattson) at 28, 1982. A copy of the Comission briefing on January the memorandum was also provided to Region I staff at that briefing.
A story on this memorandum appeared in'the February 8,1982 New York Times _ (also attached). There are two erroneous First, it fails 1mpressions lef t by the Times article.
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to note that in the memorandum the operator actions were compared to new, not existing, emergency procedure guide-The new guidelines lines for Westinghouse reactors.
re currently under review by the Reactor Systems Branch, They were not being used by the Ginna ope a
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-approved and implemented at operating plants.
Second, the Times _ article makes the memorandum appe'ar t
. conflict with other.statenents by NRC that the.Ginna operators Conta ct:
R. Mattson, HRR 49-27373 G204000238 820322 PDR ADOCK 05000244 i
The Commissioners '
performed well.
The memorandum says "it is premature to judge whether (a) the operator actions were correct, incorrect', or could stand improvement, and (b) whether the (new) emergency guidelines are correct or not."
The memorandum was speaking for the Reactor Safety staff in the Division of Systems Integration of NRR.which, at that time, had not received a copy of the actual Ginna procedures or a written chronology of the events of January 25. Obviously it was premature,then for that staff to make a judgment on the correctness of operator actions..They offered none.
Neither did they contradict the Regional Administrator's statements that were based on first-hand access to the necessary information. By failing to make this distinction, the Times article.
implies a division of. opinion between the Reactor Safety staff of the Division of Systems Integration and Regional Administrator when, in fact, none exists.
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William J. Dircks Executive Director for Operations Attachments:
1.
1/28/82 Staff Memo 2.
2/8/82 Times Article e
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UNITED STATES
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.)AN 28 &..
MEMORANDUM FOR:
Roger Mattson, Director, Division of Systes Integration FROM:
Themis P. Speis, Assistant Director for Reactor Safety, DSI
SUBJECT:
PRELIMINARY EVALUATION OF OPERATOR ACTIONS FOR GINNA SG TUBE RUPTURE EVENT Based on the chronological listing of the Ginna events you provided us on 1/26/82, which we understand was provided by R. Starostecki of Region I, I have asked Jukka Laaksonen and Brian Sheron to compare the operator actions and plant response to the Westinghous'e. Emergency Operator Guidelines for Steam Generator Tube Rupture l
Ev ents. This comparison is provided in Enclosure 1.
These guidelines are called EDI-0 and.EDI-3. We used the latest version presently under review by the staff as part of TMI Action Plan Iten I.C.1.
However, the technical guidance is generally the same as the earlier versions the staff reviewed and approved for the pilot monitoring program for NTOLS. However, we do' not know if the energency procedures ~ ~~'
in place at Ginna at the time of the accident were derived from or were consistent with these guidelines.
Our preliminary coriclusion is that the operator acted properly in using the PORY to depressurize the RCS to the pressure of the faulted steam generator. SI termina tion was also accomplished consistent with the guidelines, although it may have
'been' delayed longer than necessary and resulted in a brief discharge of'the "B" (faulted) generator to the atmosphere. The fact that the PORY stuck open complicated the scenario by producing a rapid depressurization which led to flashing of the upper head fluid. We also note that the SI pump was i estarted at 11:15.a.m., which re sulted in a second lifting of the "B" generator safety valves. The reason for this action is unclear.
. One observation we drew from this action is that operators appear to be very hesi-tant to terminate HPI when they are allowed to, or even are supposed to. We point this out since, for the pressurized thermal shock. issue, the industry has tried to
- P convince us that operators wouTd always terminate HPI before the primary system was, l
unacceptably repressurized.
Another observation is that the operators tr'ipped the..RCPs according to present instructions, and restarted the A Loop RCP when allowed. A discussion on the RCP trip ' criteria is provided in Enclosure 2.
'A number of other preliminary observations were made of the Ginna event which, I believe, warrant further investigation..These are:
-CONTACT:- J. Laaksonen, x29400 B. Sheron, x27626 e 82.D;;t&&p+-
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- Rooer l'.attson M 2 B 19E'-
1.
Stratification of the faulted steam oenerator - The faulted steam generator was being cooled ano depressurized by tne primary system. There was:some" evidence, based on thermal-hydraulic conditions of the system, that signifi-cant stratification of the secondary coolant in the faulted generator ' occurred.
This resulted in the water in close proximity to the tubes cooling down, but leaving a layer of hot water "in,sulating" the steam in the steam dome from the cold water.
Thus, depressurization of the faulted generator seemed to proceed slower than expected.
It is not yet clear what safety significance.is as-sociated with this phenomenon.
2.- Additional coolant system failures
~
a)
Leak in "A" Loop SG If a leak also developed in the "A" loop steam generator, then primary coolant would continue to leak to the secondary, unless both SGs were isolated.
Decay heat removal would then 'need to be accomplished liy
" feed and bleed" (HPI for coolant addition; PORY for coclant discharge).
(Westinghouse, in their latest guidelines, recommends cooldown using the steam generator with the lowest level and probably the smallest leakagd.)
b) Stuck-open secondary side safety / relief valve A stuck-open secondary side safety / relief valv.e in the faulted genera' tor
. would produce a direct path for primary coolant to enter the atmosphere.
Moreover, the present emergency procedures probably do not address this scenario., The primary coolant would have to be made up with HPI water.
If the leak was not stopped, or additional cooling water supplies wera not made available, then eventually all of the HPI water from the refuel-ing water storage tank would be exhausted and a net loss of primary coolant would occur. Without corrective action, core uncovery would eventually occur.
3.
Loss of steam-jet air' ejectors The loss of the steam jet air ejectors due to low "A"-loop SG pressure.(< 150 psi) produced a loss'-of-condenser vacuum and required decay heat removal by steaming to the atmosphere.
Reasons why the A loop pressure was dropped so low, the reasons why the air ejectors were lost, and the significance of this in the course of the accident will have to be addresse3.
I believe.it.fs premature to judge whether (a) the operator actionc were corr.ect, incorrect' or could stand improvement,.and (b) whether the-emergency guidelines are correct or not! This is'.because they are designed to cover a multitude of
. scenarios,-in.which the Ginna accident was just one.
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r Roger Mattson #*n 2 d 8" E'-
RSB has been addressing a similar scenario ' n response to an AE00 concern.
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will continue to investigate the above arear of concern, as well as any others brought to our attention, and recommend action as appropriate.
Z M;Li "hemis P. Speis, Assistant Director
'for Reactor Safety, Divisich of Systems Integration
Enclosures:
As stated cc: H. Denton E. Case D. Eisenhut
'S. Hanauer R. Vollmer H. Thompson C. Mich31 son G. Liinas G. Holahan T. Ippolito ACRS (1)
L. Rubenstein W. Houston O
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REi1 ARKS-f'Th'es2 remarks are based on the generic W Ererg;ncy Op;rator Guidelines fcr Steaat Generaldr.
TIME EVENT /0PERATOR ACTION i
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. Tube Ru;iture Events presently under staff review..
.a.. =...
b 9:25 SG tube rupture in SG B,.' indicated y charging pump high speed, SG high level, steam /feedwater mismatch, air ejector I
radiation
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-Trip on. low pressure, SI signal on Reactor Tripping 'on low pressure and fast depressur-l Iow pressure, containment isolation result-ization indicate a severe rupture; Westinghouse reso
.ing.ini oss of instrument air and normal for a double-ended one tube rupture do not show l
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charging flow and RCP seal flow as large depressurization.
9:29' Pressurizer level off-scale low, RC pressure '
RCP. trip cciteria are met, the operator should tElp decreased to 1200 psi the pumps manually 9:33 Operator trips RCP's Correct action, delayed 4 minutes 9:40 Operator closes main stea.m isolation valve The event has been diagnosed and the faulted SG has been identified from increasing level and pressure at the faulted SG' 8 and.the first correctiv'e action is as told in the guidelines.
It is unclear if the auxiliary FW to tt faulted SG is stopped and if all isolation valves ir tha steams lines are closed.
9:46 RC pressure 1200 pst RC average. temperature Redctor coolant temperature indicates that fast 475*F cooling in compliance with the guidelines is in progress.
9:53
.Godd SG A at 540 osi, level.76%, steam The status of SG A and the loop A is as told in the dumped to the condenser, natdral circulation guidelines. The initial cooldownof the RCS is almos in. loop A, faulted SG B at 826 psi, level completed. At this stage the operator should start Bg%
depressurizing the RCS to the SG B pressure to stog the tube leakage.
Because the RCih are o he POI should be used.
Ilowever, PORY is not aval be because of isolated 1nstrument air system.
Faul ted
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' SG pressJre is surprisingly low and indicates lack proper isolation.
'9:57 Operator reseh SI, pumps still on, instrument At this stage; the operator is able to depressurize etr re-established the RCS, but he does not do that. Thus, the leak
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continues and SI is needed to keep the RCS inventor
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10:04 Charging pumps restarted, RCS at 1300 psig, pressurizer level 18%,
Rastart of charging pumps b'jf;re stepping tiv-i.u i.
SG B level 100% narrcw range, 400 inche.
1cak is nst in corr'plianco.wlth tha guidelin2s ane[
s 1cak increases b::cause the charging pumps Ir.crea.
. ide range i
w RCS pressure.
SG B filling up is a cnnsequenki a failure to depressurize. the RCS.
10:07 Operator opens the FORY manually The action is taken somewhat too late 10:08 Operator opens the PORY manually for a second If the PORY is opened on time there is
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time, PORV sticks open multiple PORY operation.
no rea 10:09 PORV block valve shut,RCS pressure 800 psi, Correct action to shut block valve.
Drop in tu
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p essurizer level offscale high.
pressure to 800 psi flashes hot liquid to steam ir RPV upper head, pushing water into pressurizer 10:10 SI pumps increase the pressure to 1300 psi Af ter having closed PORV the operator is instru. cit.
see if the system repressurizes.
If, pressure in-creases 200 psi, then the SI can be turned nfr.
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to stuck open PORV, the ability of the operator te follow this guidanch is questionable.
10:18 T
. incore = 458'F Adequate subcnoling nunusperic res ier va ve.un u
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B isolated manually Precaution against opening 10:30 TPV head = 525'F (p sat =850 psi) flot clear if upper head bubble cundensed eis.
Temperature supports existence or bubble 10:40 Safety valve of the S[ B blows. ope.rator Safety valve blowing is a direct result of secuEin secures the SI to reset the SI the SI about 30 minutes too late 10:42 RCS at 800 psi There is obviously almostno steam bubble in the in ssurizer and the pressure decreases drastically whc some water is drained from the system.
Pressure o
stabilizes to a value where the water in the uppe head starts to flash.
It is unclearhow primary u
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is lost because the faulted SG B ~should be at abou safety valve set pressure (above 1000 psi).
10:50 RC pump' seal return relief valve lift and dis-Reason for relief valve opening unclear, possil.ly charge to the PRT
. related to the containment isolation E'
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IIME lkVENT/0PERATOR ACTI0tl.
REMARKS
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"10':57 PRT rupture disc blows, total loss PRT. overpressure results from combined seal. return alief valve operation of coolant S 8000 gal L
11:15 SI pudp* started,SG B safety valve Reason for SI start unknown, SI re-initial.lon lifts at 1035 psi (set pressure criteria are not met (loss of sub -cool ini, 1085 psi) pressurizer level <20%); when did the safety va resst?
11:15 T cold on operable loop was higher than core Reason unknown exit TC's 11:22 RPV head at 525'F, 97'F subcooling, pressure 1035 psi 11:29 r
RCP in loop A restarted, core exit and The guidelines tell to start at least one i
upper head thermocouples equalize at after the initial plant stabilization and SI 450'T termination, restart may be delayed because oT problem with seal in.lection 12:00-Pressurizer level 80%
- 12:05 Normal letdown established Nonnal letdown /cliarging is' told to be establisti as the next step af ter Si termination 12:30 Slow cooldown started, RCS'at 923 psi 1
e 6:40 Level' in SG B re-established SG cooled SG B depressurization is obviously no problem' by adding cold auxiliary, feedwater because there is not much steam lef t a
7:05 RHR initiated, RCS at 280 psi, 330*F l
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ENCLO!L'kE 2.
. There are a number of questions that can be raised by the pump trip issue.
The RCP trip criterion for W plants is that.the operators shoilld trip all RCPs if SI has actuated and the primary system pressure has dropped below a specified value.
This value (or rather the method by which this value is obtained) was worked.out and mutually agreed upon between W and the staff during the B&O Task Force after the accident at TMI.
Essentially, the value is the set pressure of the secondary safety valve plus adders to account for pressure drops back to the pressure gauge in the control room.
Typical
. values are expected to be between 1350 and 1450 psi.
A more detailed dis-cussion of this pressure setpoint derivation is provided in Section 7.2.2 of NURELOS23 (Attachment 1).
It is noted that the charging flow at Ginna is isolated on an ECCS signal. Since seal injection flow to the RCPs is from the charging flow, it too is lost during an ECCS signal.
Thus the pumps would be required to be tripped following the ECCS signal, regardless of the pump trip criteria.
For any steam generator tube rupture which depressurizes the primary system to below the RCP trip pressure, there would be a need for the ~ operator to use the PORV to aid depressurization, since the primary system pressure will 1.
stabilize slighty above the fMited steam generator secondary pressure.
For a smaller leak, the RCPS would not be tripped 'imediately. and the sprays would remain availaole to aid in the depressurization.
For..CE and B&W plants, the RCP trip criterionis on low pressure SI actuation (around 1600-1750 psi).
Th'us earli'er RCP 1 rip would be expecte'd for:.both these' plants.
Both CE and B&W were asked informally.to adapt the W low pressure criterion, but.both declined.'
Section 6.0 of NUREG-0623 recomended that the industry develop RCP trip criteri
~
which tainimized RCP trip for non-LOCA transients (see Attachment 2) and also recomended that procedures,and training be initiated for handling non-LOCA events which produce ECCS ectuation anu pump trip; including instructions for:
'a) tripping RCPs:
b) monitoring a}nd initiating natural circulation; c) pressure control without pressurizer spray;
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d) HPI termination;
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e) RCP restart criteria.
. provides section 7.2.5 of NUREG-0623.
The reasons for requiring RCP trip are as follows:
For certain small breaks in the p.rimary system, continuing RCP operation e
will " pump" water out of the break, and produce a. greater coolant invenfory loss than if the pumps were tripped.
For any, small break in the primary system; including steam generator tube e
rupture, the coolant will become two-pha'se and could evolve to a.
significant void
- fraction.
We are no.t' aware of any RCP's that have been designed to operate for extended periods of time in a highly voided system.
Continued operation in a highly voided system could result in excessive vibration and possible seal failure, or worse.
During the initial phase of many transients and accidents, the symptoms
,e may resemble those of a srhall break or a steam generator tube rupture.
- r.
Early RCP. trip with restart instruction is considered the most prudent course of action.
with a small break and Anal $ sis models which predict system performance' s
the RCP's operational still have large uncertainties when the system voto Thus we do not have a high conf,idence that pump failuie.
fractions are high.
during high void conditions does not lead to unacceptable core uncovery.
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Short-Te." Recuirements
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The f ol10.1..g ooscribe the short-ter.. recuirements for pump trip for each of the reactor vendors.
Control F.oce Ooerators 7.2.1 JE Bulletins79-05C and 79-06C (item 1.B) require that two licensed operators be in the control room at all times (three for a dual control room) and that one of the two operators be designated to trip the reactor coolant pu ps should the facility undergo a transient which results in a The safety injection actuation signal due to low primary system pressure.
designated operator may perform any normal or routine control room duties at all other times.
The licensee should confirm that an operator is designated to perform this action on each shift.
l 7.2.2 Westinchouse-Desioned Plants For the sh:rt-term, tne staff has adopted the following position for manual pump trip requirements on Westir.gnouse-designed plants.
f Staff Position on Pumo Trip for Westinohouse Plants We require that the reactor coolant pumps be tripped at a system pressure determined in the following manner:.
l Secondary System Pressure - Based on the number and size of the l
(1) secondary system safety valves, the secondary pressure will be l
established by determining the pressure setpoint for that v.alve in which the calculated steam relief is less than 60 percent of the
?
I valve's relief rating.
If the calculated relief is grea'ter than 60 percent of the rated capacity, then the,next highest pressure setpoint should be used.
Primary to Secondary Pressure Difference - To account for' the pressure (2) gradient needed for. heat removal, pressure drop between the steam generator and safety valves, uncertainty of the safety valve setpoint, pressure drop from steam generator to measurement loc.ation, etc., the
..j primary pressure for RCP trip should be the secondary pr
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are 100 psi.or less.
If the adjustment are determined to be greater i
j than 100 psi, the larger value should be used.
Instrument inaccuracies appropriate to that time in the loss-of-coolant; (3) accident should be added to the primary pressdr.e established in (2),
The resulting pressure is the indicated pressure at which the above.
l l.
operator should trip the RCPs.
~
Combustion Enoineerino-Desianed Plants tombustion Engineer 1Tig has recommenced that reactor coolant pump trip'be 7.2.3 manually initiated by the oper: tor on receipt of. reactor trip 'and safety injection actuation signals.
Combustion Engineering is also evaluating the capability of their plants to accomodate a pump trip o'n reactor trip and a lower system pressure by a method similar to that established for Westinghouse as specified in Section 7.2.2, above.
The staff will accept the pump trip based on reactor trip and SI acteation fcr the short-term, since SI actuation pressure is approximately 1550 to, 1
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LCU. c2n result in a greater c:n irvertory less from the syr* c-t..cn if :ne pumps were trippec.
The ability to correctly represent the thermal-hydraulic behavior in key comporients within the primary system during a small break LOCA (2)
Moraover, it with the reactor coclant pumps running is questionable.
is unclear at this time which models clearly result in conservative, This is substantiated by the variety of bounding calculations:dif ferent models used to represent the various prima in vendor analyses and the differences in the limiting small break It is our conclusion that this uncertainty in thermal-hydraulic modeling presently precludes the use of these models for analyses.
quantitative determination of smalT break system behavior.w coolant pumps running. substantiate allowable modes of pump operation LOCAs.
It is our conclusion thet for the pumps running case, insufficient integral system experimental data presently exists to substantiate (3)
Moreover, we do not the quantiative results of the analysis codes.believe any pro with that necessary for short-term resolution, which includes the addition of equipment necessary to assure automatic tripping of the coolant pumps for small break LOCAs.
From items (2) a'nd (3), above, we find th'at tripping all of the reactor coolant pumps during small break LOCA (4) equipment.that is safety grade to the extent'possible.
i, The impact of an early pump trip on non-LOCA transients is not pr However, tripping the reactor (5) to lead to unacceptable consequences.
o coolant pumps for non-LOCA transients can aggravate the consequences I
of these transients and ex* tend the time r'equired to bring the plan into controlled shutdown condition. reactor coolant poings cur r
i potential for interruption of natural circulation due to steam forma-i tion in the coolant loops.
Therefore, we conclude that the criteria and te uirements for rea'ct
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coolan_t_ pump ~tTiBo' be esl'ab'llThed"fRii~itF l
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reactor coolant pump trip for non-LOCA'transi~iiC The staff recognizes the potential desirability of running the reactor.
coolant. pumps to provide forced circulation during small break LOCA (6) and we encourage the continued exploration by the industry of.mean by which this could be accomplished.. capacity or two pum are a step in this direction.
We will require verification of small break models with the pu=ps -
In running against appropriate integral systems experimental tests.
(7) particular, we will require that the NR vendors 2nd.fue1 suppliers
- J
1500 psi; ihr.: r :..u as c:mpreo to 51.ctenica pressufes of about 3B00 to M. r 5ir 1
- ..es tir.ph:.se clar. 5.
I-is exoec ed that the pressure used fcr ; u ; tri
- t. '..':r-i.g.:.sc wir 'all
- Tr:ximately in the rangs of
,the saf ety injection actLation presserc-fer bcth CE and B&W plants.
Embcock and Wilcox-Casicnad Plants 7.2.4 Eaccock & Wilcox is also recom ending that for the short term, pump trip be_ manual.ly initiated on automatic actuation en low pressure of th{ safety In addition,' Babcock & Wilecx and their plant owners '
injection system.
are examir.ing the pcssibility of a shor -serm manual trip requirement based on subcooling rather than automatic SI actuation on low pressure
'The staff agrees in principle with this a proach, but final approval only.
F must wait until the details of such a method have been formally submitted and evaluated.
The staff finds the present short-term requirement for manual trip on 5
automatic SI actuation on low pressure acceptable.
E&W SI actuat. ion setpoints are between 1500 psig and 1650 psig and are considered consistent i
with the setpoints at which the pumps would be tripped for both Westinghouse and Combustion Engineering. plants.
Trainino Guidelines and Emeroency Procedures 7.2.5 IE Bu11etins 7S-05C and 79-06C (items 3 and 4) requested the' Westinghouse,.
Combustion Engineering, and Eabcock & Wilcox plant licensees to:
(1) Develop new guidelines for LOCA and non-LOCA events based on LOCA -
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analyses and RCP trip requirements, and e
Revise emergency procedures.and train all licensed operators and 9
(2) senior reactor operators based on these new guidelines. '
P In general, the licensees have identified guid' lines, procedures, and training for loss of, coolant events in their responses to these items.
This effort on LOCA events was already in progress at the time the bulletin was issued.
' Because of the potential for initiating ECCS by other depressurization i
events such as overcooling because of a malfunction in the secondary system, the operator would have to trip the reactor coolant pumps before g
As a result,
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he could make a determination about what event is occurring.
- l we require that the licensee have procedures and operator training to i
handle non-LOCA. events which may also have ECCS actuation and reactor l
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coolant pump trip.
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The procedures for these non-LOCA events should include ins (ructions o t
flI tripping the reactor coolant pumps, monitoring and initiating natural circulation, pressure control without the pressurizer spray, HPI termina,-
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The licensees tion criteria, and reactor coolant pump restart criteria.
should confirm that these procedures for non-LOCA events are in place and l
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l the operators have been trained in their implementation.
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THE NEW YORK TIMES. MONDAY, FEBRUARY 8. Ht2 s'
$. 5,y MAT"DfEW L.,WA!.D ",'. ".
'seems tc heve made their job mon dffi.
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,'" WASHINGTON, Fsb. 7-Radioacuvity wa-
- lna's tfro~~stearn generators. Ai G nna For example," when the leak began
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j released into the atmosphere during the recen' and at 48 other plants of its type amund and the rencor sh:t down, anctber sys.
nuclear acddent at the Robert E. Ginna pizrv,
the coun~y, radioacdve water is cem. tem auta=atically began destag valves i i because operators were "too' late"in tumiry, lated around the bot uranium cort, and and shutting down parts of the plant can.;
- cff emergency pu=ps,accertiing to a prelim 3 then pumped through thousands of nar, sMered nonessenual in an. emergency;
[ nary evajustion bya staff member of.the N. u ;
row tubes, when it Efves off its ben. hof those pans pmWded --
. dear Regulatory Ccemission.
~!.G O.:tside the tubes, dean water is bcGed airto operata a valve that can be used to A second release occurred 35 minutes later '
"into steam, which is used to run a tur. nduce be pnssun of W rascacun because operators restarted the pumps for bme..
water, a useful step ti there is a leak.
" unknown" reasons,the study found.
The radioacdvs water, at tempers.
~ Shut Down Pr.!rof Pumps t ses ver degrees Fahnnhest, is Aho as a safety step, the operators Chr.o.no)o378f Accidest under pressure to pavent boiling. shut down a pair of pumps Qu, among
", The evaluation said it was "prematurt" 14 i ' hat ha ed at Ginna is that one of mher functions, can be used to cool bjudge whetheractiocs by o;iifators wert "cor.
W burst, allowing the pmssur* waterin the device
- i. rect, incorrect or could stand improvement."
ized radioacuve waterto rtinto Be andthereby werthepressure..
1,.E,owever, its author and other officials.of the clean st cas De operators restarted the air !"ine to commissiczicalciahed to'.ay'that the opera.
gWrm W weerW mstam the valve that relieves on the tors had hardled the accident well. A spokes.
- man for the 1tochester Gas and Electric Com.
.in me steam generatcu.% pnssum radioac:tive water, but not the -
valve unt!!(2 minutes afterthe E pany, Judy Houston, sald that t!ie utility had fch omsid thed began. "De action is taken somewhat
,g3 not seen the'd6cument and therefore could not tainmem buDeng'.
too lata." remarks the evaluation's run.
cornment onit,.c
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Certain safety st it is clear in ret. ning commentary on 2e CJE The evalustion s'so Nnts to problems in the respect, aggravate [t$e pmblem. Auto.Thoalw was men closed,butme pas.
safity systems of.the plant, a Westinghouse de-maticaDy, the reactor's i:nntrol system sunwas suu tcohigh.,;. g sign in common use', which made the a'ccident '
sensed the loss of water in the radioac.
One minute,later,.the operators
' U ardeY to contro!. Qne major plant'gompoor,tr.
ve loop and turned on emergency opened the The analysis remarks thatif valve a second time, and it h
' used to' regulate the pressure in t!ie reactor was" pstoidd water.In mdaycases,this stucs
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' disabled by a combination of safety actions, ac/!
a==my step to insurt that the the had been opened a timac
' cording to a chronology of the acddent that it.
cort remains covered with. water.If the menwouldhanbeennoreasutoopen
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'coreis wmM, as ha mThree itasecond%
.,g ny..
E."artof tidevaluaticaz.
74.' Qg..q'#-The evaluaticm, whle. lits stamp 7d '.*d'ralt," l lMllelsland,the resul beatwddam.
The analysis says,that the of
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' was written by Themis Ps Spels, assistant df i ageit. 4 - % M~r ifa &. -
events in'the Ginna accident raises the -
sector for reactor safety of the Division of Sys. l
'P
I M on!,odger $," possibattyof omercombinations,cQ,
"tems Integration..The division is studying the !
!$thhhee M!ie Island case,the ac. " " ' * * * * "
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problem of automatic reactor systems thatin.
l cident was compounded by the opera. / For 2e chrondogy raket
.teract in tmexpectedyays and cause pmblems
- tors' dedslon to turn off the emergency me quesdon whypnssumdappedso.,
ps were left..sharplyfyhen, es, sat valw m te
, in emergendes.'s me.'.r.F'-4 :7 pr'efcy--?
on ong. At G!nna, the
- The acddent on Jai 2. 25 at de lant:16 m!!es er. Turning them
- may have steam gerianer 2e Erst 13 5een delayedlongerthan na=wy and lease to Be atmosphen, implying Gat :
northeist of Rochester, restut in the secon.
. tmiled release of a small amount of radioactivi.
resultedin a brie!
" to the att thnalumayhan stayed open sustaly, I ty,although such releases have been avoided mosphers, the analysis
- d. The valve. lon man The that allowed the radioact'.ve steam to es. he ts g
ys,,says E.during simild acddents at other plants.:
tirae as a " direct M what amount o radia'.fon released wai not danger.
cape opened the first,down the.
lcus, ac.:c{tding to'pikt'and com nissiod offi.
would ppen c a valve stockin
, result" of shutting ps
. dais.
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about 30 minutes too late,,.acco tkmodum andpmduced"adimet imarycodut w esettbem.
j Accortfing to, Mr. Spels's memo, the G!rma, tothe when the meisphere." Accortilag to the analysis, accident raises,the specter of f ar me. e_seri_ous--
[hw,ere rwh!chsaidth'ereasons orthis probably do not address mis scenario."
tarted a to Be "The present emergency proceduns "acBdentscsome of which are not anticipated by current emergency procedures. For e.xample, acham'w h "
In such' a dtuadon, smless 'Ge leak remarks secompanpng the chrtmology of j'
events ic' ply that a crue!s.! safety valve may Ronald C. Haynes, a coWssien cfS. could be stopped er a new supply of-I cial who was at the scene on the day cf water could be fo=2d, the emergency tmefly have swck cpen, and 11 it had stayed the acdde=t, said that even though re-pu=ps would c=ntinue to add water, epen, sm. mg to the analysis,it is net clear starting the pumps resulted in a release wh!ch would be pt:mped through the how dsmage to the reacter's core would have of radaoactivity, it was a "conserva. burst tube er other leak, tu=ed into been prevented.
tive" step, because the operaters steam and released to the e=virec=ent, in addition, the hesitancy of the cperat=s to thought 1t would reduce the possibi!!ty of until the erne:Eency system had no
. ~
"tm off ct.-tain saf ety pu=ps in this case raises cereca= age. t.
mere water left, and "a net 1::ss of ;:ri.
he pcesibility that in tumre cases, where such
,.The ana.tysis said that "cperaters ap. mary coolant would occur. Without cota a thm.cff mould be men impertant, the pu=ps pearto be very hesitant" to tum c!! the recuve actics, cere,uncovery would wculd be left en and pressure w ould rise so high pu=ps "when they are allowed to, or even:: ally tw."
4.
cat the reacter ves.sel itse.!!, which bcId the even supposed in." ~....
The result of "cc:t uncovery"is core i
fuel. night be cracked: 1 In additics to these acticas by the damage, the most serious kind cf sed.
Mr. Sp-Is's memcrandum steps short cf cperaters, the des!gn c! the reactor dent.
teferri.g to the actions of the cperaters as er.
I re s. licwever,his chrtnolegy of the accident l
no es dedsi:rts made by eperatn.s thatztsu!ted l
in the leak rua.ning faster, and in a safety va!ve
" blowing" twitt and allomtg rad:oactive ter::ta ventinto the attaosphere.
The a t:'. dent bege.. at 9::.5 A.M. in ene c! Gin..
- M*NETY.CCVENTH CONGRE35 CHinLES CONKUM P
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ASSOC 11TE FFE1AECTOR srr w sYr so. was.
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COMMITTEE ON INTTPIOR AND INSUI.AR AFFAIRS Ano couwsco n
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February 5, 1982
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The Honorable Nunzio Palladino Chairman United States Nuclear. Regulatory Commission Washington, D.C.
20555
Dear Mr. Chairman:
As a follow-on to the February 4 briefing, I would appreciate the Commission's answers to the following questions in addition to the information requested by
!!r. Markey.
i 1.
What is the primary significance of the Ginna incident?
2.
What was the leak rate through the break as a function of time?
r; 3.
What triggered the steam generator tube rupture?
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i 4.
Had'there been indications of leaking steam generator tubes prior to the rupture on January 25?
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i 5.
What was the caase of the PORV's apparent failure to closje? Does the apparent failure of the PORV to close cause EuTt ab'out the adequacy of the industry's program to test such valves?
6.
What would the course of the incident have been had the PORV block valve failed t.o close partially or fully following failure of the PORV to close fully?
~
7.
Did the procedure for responding to a steam generator tube rupture contain instructions for actions to be taken in response to development of a steam bubble in the' reactor pressure vessel?
8.
Was there a need during the incident to take actions not specified in the plant's written operating and emergency procedures?
Were the emergency procedures in place at Ginna consistent with Westinghouse guidelines as discussed in the January 28 memorandum from Mr. Spels to Dr. Mattson?
9.
Had a water level measuring device been available, would it have assisted the operators in determining the size of the steam bubble in the pressure vessel and otherwise in bringing the plant to a stable condition?
L(h h A4L168.146 l
i Chairman Palladino Februcry 2, 1982 10.
What consideration has been given the potential for radioactivity escaping PWRs via a path including breaks in steam generator tubes and a stuck open safety valve.
11.
Is it generally agreed that if a leak had developed in p
both steam generators, the operators would have been able to institute the " feed and bleed" process described in Mr. Speis ' January 28 memorandum.
(
12.
How many steam generator tube ruptures per year of the Ginna magnitude or greater do you expect?
13.
What is the likelihood of several steam tube ruptures occurring at one time?
What is the maximum number of I
simultaneous or near simultaneous steam generator tube ruptures.that are considered design basis accidents following which the a can be brought to a safe shutdown condition by following plant operating and emergency I
procedures?
14.
Did WASH 1400 or more recent risk assessments determine I
the probability of occurrence of events in which one or more steam generator tube failures are followed by various combinations of PORV, block valve, and safety valve i
failures?
15.
How long did it take to reach cold shutdown?
Is this a period longer than desireable?
What was the reason for the period being longer than normal?
What kinds of malfunctions during the extended cooldown period might have led to a significant release of radioactivity to the environment?
16.
Did any part of the reactor p.tessure vessel cool at a rate in excess of that stipulated in the plant technical specifications?
17.
Was there a capability at Ginna to remotely vent the reactor pressure vessel high points?
Does the Commission i
believe that conditions might develop in PWRs calling for the use of remotely controlled valves for the purose of venting steam?
18.
At any point during the Ginna ever.P., did the steam generator containing the ruptured tube control the primary system pressure?
Are operators at Ginna and other PWRs
'ttaIned with respect to actions to be taken when a steam generator controls primary system pressure?
l Sincerely, g
i l
S Chairman
.