ML19224C400
| ML19224C400 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/12/1979 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Case E Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7907020279 | |
| Download: ML19224C400 (6) | |
Text
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ME. 0RMTUM FOR:
F. (i. f ace. Perub Direct or Of f we ef ti icar !:ca !er $;e.; J a t i en FROM:
D.F.Ross,DeputyDirectorCQ>yf Division of Project Managemefd
SUBJECT:
SUMMARY
OF MEETING WITH WESTINGHOUSE - CORRECTIVE ACTIONS FOR WESTINGHOUSE NSSS PLANTS AS A RESULT OF THREE MILE ISLAND UNIT 2 INCIDENT On April ll,1979, the NRC staff met with representatives of Westinghouse Electric Corporation (W) in Bethesda, Maryland, to discuss short ierm corrective actions to '.e implemented at Westinghouse pressurized water reactors (PWR) as a result of the incident at Three Mile Island Unit 2.
Several W PWR licensees were in attendance. A list of attendees is attached (Enclosure 1).
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The meeting opened with an overview of the events at Three Mile Island Unit 2 (TMI-2) which require immediate attention af all opera'.ing PWR's as these events ara perceived by the staff in light of information available at this time. These events are identified as Items 1 thru 12 in the NRC Office of Insoection and Enforceaent (0I&E) Bulletin 79-05A of April 5,1979 (Enclosure 2). The staff specifically noted that the responsibility for development of corrective actions for these items rests with W and the utilities. The correct;ve actions that are needed are speFific instructions to be issued immediately to licensees of W PWR's. These corrective measures wil' be reviewed by the NRC staff and issued by means of an 01&E Bulletin.
'{ representatives then presented a summary of tk activities which they have initiated since the TMI-2 incident to prevent the occurrence of a similar incident at a W facility.
Since April 1,1979, W has been working with its customers on this issue, and on April 5,1979, a meeting was held between W and its customers to discuss the potential for the occurrence of a TMI-2 type incident at W facilitiec.
Since then, W has been conducting additional studies concerniag specific plant concerns regarding the TMI-2 incident and has conduc ted some computer analysis of the incident.
W has also asked individual utilities to compile plant specific information which may bear on the probability of occurrence of mitigation of a TMI-2 type incident. W representatives stated that the efforts undensay with their customers covers all the items identified in IE Bulletin 79-05A and some additional areas of review.
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E. G. Case
-Z-AFRIL 1 2 1973 h[ then discussed the response of a typ..al 4-loop (four reactor coolant system cooling loops) PWR to a loss of feedwater (to the steam generators)
The transient response "eported in individual plant Safety transient.
Analysis Reports (SAR) is more conservative than the actual response For actual transients, experienced at h[ facilities for loss of feedwater.
the large steam generator secondary-side inventory provides a buffer between secondary (steam side) transients and primary (reactor coolant system) resporise to the transients. h[ is still investigating, but as of this date, they are not aware of any loss of'feedwater leading to a primary system pressure increase that caused a pressurizer power Therefore, a stuck open PORV operated relief valve (PORV) to open.
similar to that experienced at TMI-2 should not occur for a loss of
if no credit is taken in the analysis for non-safety grade plant control system, PORV lift will occur; and there exist other transients which can lead to a PORV lift (and to the potential for a stuck open PORV).
Because it is not impossible to preclude PORV lif t and the potential for a stuck open PORV, W performed computer analyses using conservative assumptions to determine the response of a typical 4-loop PWR to a Using the W-FLASH code and 10 CFR 50, Appendix K s tuck open PORV.
assumptions, W analyzed a 24" dia. Loss of Coolant Accident (LOCA) break in the vapor space of the pressurizer. This break size is similar to the size of LOCA caused by a stuck open PORV. h[ also assumed the steam generators were isolated (main steam isolation valves are shut) and no charging flow makeup to the reactor ccolant systems is in progress.
Three cases were analyzed:
Case 1 [with auxiliary feed system (AFS) flow to steam generator and with safety injection (SI)']
The reactor core remains flooded with cooling water Results:
a.
throughout the duratioc of the anaiysis (Approx. 4000 sec.)
and the parameters indicate that no uncovering of the core would occur thereafter.
The pressurizer steam-water mixture level increases and b.
stabilizes at about a 2/3.
Case 2 [with no AFS and with SI]
Results:
a.
Same as Case 1, a. and b.
b.
2/3 of steam generator level is still present at 4000 sec.
Reactor coolant system pressure approaches 1100 psi which c.
corresponds to the temperature in the steam generators with safety valves lifting.
E. G. Case APRIL 1 t 1373 Case 3 [with AFS and no SI (i.e., no water makeup to the reac tor coolant system at all))
Resul ts :
a.
The reactor core would start to uncover at about 2100 sec.
Additional analysis is being done by W for Case 3 withou the steam generators isolated.
And a comparison of Cases 1 and 2 ndicate that the results are not very sensitive to AFS initiation for the time periods of the analysis.
W discussed the signals wFich initiate SI. Analyses which they submitted previously (Zion Sta' tion and RESAR-3 dockets) shcw that a srall LOCA in the pressurizer steam space may not result ir SI initiation because the pressurizer level may not decrease. A coincident pressurizer lcw level (Lp) and low pressure (Pp) is needed for SI actuation.
But their analysis of containment builo:'g pressure following this LOCA shows that SI would be initiated by containment pressure high (no.1) indication setpoint which is set at about 10% of containment design pressure at about 1600 sec. At 1600 seconds, reactor core fuel surface temperature would be at the same temperature as the reactor coolant system coolant which is saturation temperature for 1100 psi. This is far below the temperature necessary for core damage. To provide additional assurance that SI initiates and prevents the core from becoming uncovered, in addition to considering the high containment pressure setpoint SI actuation signal, W has instructed its customers that SI should be manually initiated if Pp decreases to the low Pp setpoint regardless of Lp readina.
W is still evaluating the question of when to manually chut off SI following its activation.
The concerns are (1) that the SI system would fill the reactor coolant system completely and thus increase the chances of an overpressure transient which could overpressurize the reactor coolant system or (2) that the operator would shut off Si based on an erroneous pressurizer level and thus increase the chances i presented a logic " tree" of a TMI-2 type incident (core uncovery).
l that an operator could use to detennine if SI should or should not be shut off following events which lead to SI and low or toiling pressurizer pressure and/or level.
W agreed that a bulletin similar to Bulletin 79-05A should be sent to Its customers, but additional clarification of the need to shut off SI to prevent overpressure as discussed above should be included. W noted that the bulletin provision regarding containment isolation reset is not applicable to its plants because containment isolition valves do not open following an SI reset (as occurred at IMI-2) unless the operator deliberately opens the isolation valves.
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E. G. Case APRIL 1: 0 79 Following the h[ presentation, the staff discussed the followup action to be taken in light of the h[ information.
A belletin will probably be -issued to W facilities in the next few days. The bulletin will be essentially the same as 01&E Bulletin 79-05A but additional information will be included to determine plant specific corrective measures dealing with:
1.
Manual shutoff of SI, 2.
Management checking of safety system operability status, Possible elimination of Lp as an SI initiation s'qnal by placing it 3.
in a " tripped" state, 4.
Possible requirement for containment isolation on high radiation signal for all plants.
The bulletin will state that our best information shows that, under certain transient and/or accident conditions, a level may be present in the pressurizer simultaneously with a decreasing primary system pressure.
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DenwoodF.Rcss,DepukyDirector' Division of Project Management
Enclosures:
As stated cc w/ encl:
See next page 264 303 0
E. G. Case S-APRIL 1 2 1975 Distribution Docket (50-320)
NRC PDR Local PDR 00R Reading NRR Reading H. R. Denton V Stello R. Vollmer W. Russell B. Grimes T. J. Ca rter D. G. Eisenhut A. Schwencer D. L. Ziemann P. Check G. C. Lainas D. K. Davis T. A. Ippolito R. W. Reid V. Noonan G. Knighton M. Fletcher D. Brinkman Attorney, OELD R. Fraley, ACRS(16)
J. R. Buchanan TERA NRC Participants
'264 30!;
Enclosure i LIST OF ATTE.NDEES WESTINr, HOUSE MEETING 04/11/79 NRC Carolina Power & Licht R. S. Boyd, DPM D. B. Wa ters T. A. Ippolito, D0R J. J. Sheppard M. H. Fletcher, 00R R. Lobel, 00R Shaw, Pittman, Potts & Trowbridge E. G. Case, NPR F. Orr, DSS J. H. O'Neill A. Ignatonis, DSS N. C. Moseley, I&E American Electric Power Serv. Corp.
J. L. Crews, I&E Region V E. A. Reeves, 00R J. G. DelPeriro N. Anderson, DOR E. Wenzinger, 00R' Public Service Electric & Gas M. Mendonca, DOR L. B. Marsh, DDR P. A. Moeller D. Neighbors, D0R J. Wetmore, D0R L'es tinghous e T. V. Nambach, DOR A. Burger, 00R R. W. Stutter B. C. Buckley, DPM K. R. Jordan A. J. Szukiewicz, DSS V. J. Espusito J. Guibert, OCM W. J. Johnson A. Schwencer. D0R T. M. Anderson G. Zwetzig, 00R S. H. Hanauer, DSS Southern California Edison F. Schroeder, DSS L. P. Croker. DPM J. Rainsberry D. Vassallo, DPM A. Thadani, DLS G. Lainas, DOR D. F. Ross, DPM D. G. Eisenhut, D0R I
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264 305
UiilTED STATES NUCLEAR REGULATORY COMISSION
~ ~ ~ 'Enclm,ure 2~ - ~
0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 april 5, 19.79 IE Sulletin 79-05A NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Cescription of Circumstance :
Preliminary infocmation r eceived by the NRC since issuance of IE Bulleti
' 05 on April 1.1979 has identified six potential human, design ar.a mechanical failures which resulted iri the core damage and radiation releases at the Three F.lc Icland Unit 2 nuclear plant.
The informatio snd actions in this st '
. ment clarify and extend the original Sulletin ar.J transmit a peiiminary chronology of the TMI accident through the first 16 w s (Encicsure 1).
1.
At the time of the initiating event, loss of feedwater, both of the auxiliary feedwater trains were valved out of service.
2.
The pressurizer electromatic relief vaive, which opened during the initial pressure surge, failed to closa when the pressure decreased below the actuation level.
3.
Following rapid depressuriz" ion of the pressurizer, the pressurizer level indication may have lead to erroneous inferences of high level in the reactor coolant system.
The pressarizer level indication apparently led the operators to prematurely terminate high pressure injection ficw, even though substantial voids existed in the reactor coolant system.
4.
Because the containment does not isolate on high pressure injection (HPI) initiation, the highly radioactive water from the relief valve discharge was pumped out of the containment by th automatic initiation of a transfer pump.
This water entered the radioactive waste treatment system in the auxiliary building where some of it overflowed to the floor.
Outgassing from this water and discharge through the auxiliary building ventilation system and filters was the principal source of the offsite release of radioactive noble gases.
5.
Subsequently, the high pressure injection system was intermittently operated attempting to control primary coolant inventory losses through the electromatic relief valve, apparently based on pressurizer level indication.
nue to the presence of steam and/or noncondensible voids elsewhere in the reactor coolant system, this led to a further readinn in primary coolant inventory.
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