ML19246A364
| ML19246A364 | |
| 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 |
| Shared Package | |
| ML19246A360 | List: |
| References | |
| ACRS-SM-0102, ACRS-SM-102, NUDOCS 7906180637 | |
| Download: ML19246A364 (6) | |
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M NUCLEAn "ULATORY COMMISSION P
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MEMORAtiDUM FOR:
E. 3. Case, Deputy Director Office of tiuclear Reactor Regulation FRCM:
D. F. Ross, Deputy Director %
Division of Project Managemeht
SUBJECT:
SUMMARY
OF MEETING WITH WESTIrlGHOUSE - CORRECTIVE ACTIOt15 FOR WESTIt'3 HOUSE ilSSS PLAT 4TS AS A RESULT OF THREE MILE ISLAtiD UNIT 2 ItiCIDENT On April 11, 1979, the NRC staff met with representatives of Westinchouse Electric Corporation (W) in Bethesda, Maryland, to discuss short teim corrective actions to be 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 L
attached (Enclosure 1).
g.
f The meeting opened with an overview of the events at Three Mile Island Unit 2 (TMI-2) which require immediate attention by all operat.ng PWR's as these events are perceinad by the staff in light of information available at this time. Tbese events are identified as Items 1 thru 12 in the NRC Office of I spection and Enforcement (01&E) Bulletin j
79-05A of April 5, 1979-(enclosure 2). The staff specifically noted i
that the responsibility f'r development of corrective actions for these items rests with W and the utilities. The corrective actions that are needed are specific instructions to be issued imediately to licensees of W PWR's. These corrective measures will be reviewed by the NRC staff and issued by means of an 01&E Bulletin.
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I W representatives then presented a summary of the activities which they have initiated since the TMI-2 incident to prevent the occurrence of a similar incident at a W facility.
i Since April 1,1979, '[has been working with its customers on this issue, and on April 5,31979, a meeting was held between W and its customers to discuss tne potential for the occurrence of a TMI-2 type incident at W facilities.
Since then, W has been conducting additional studies concerning specific plant concerns regarding the TMI-2 incident and has conducted 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 underway with their customers covers all the items identified in IE Bulletin 7935A and some additional areas of review.
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APRIL 1 2 G73 '
E. G. Case
(( then discussed the response of a typical 4-loop (
The transient response reported in individual plant Safety Analysis Reports (SAR) is more conservative than the actual responseFor actual tra transient.
experienced at h[ facilities for loss of feedwater.
the large steam generator secondary-side inventory p system) response to the transients. h[ is still investigating, but as of this date, they are not aware of any loss of feedwater leading to a y.
primary system pressure increase that caused a pressurizer powerTherefore, a s operated relief valve (PORV) to open.similar to that experienced at T
- However, feedwater transient under normal plant ope: dting conditio system, PORV lift will occur; and there evi:t other transients which can lead to a PORV lift (and to the potential for a stuck open PORV).
Because it is not impossibl'e to preclude PORV lift and the pote for a stuck open FORV, (( ps assumptions to determine the response of a typical 4-loop P assumptions, W analyzed E 2b" dia. Loss of Coolant Accident (LOCA)Th stuck open PORV.
break in the vapor space of the pressurizer.
W also assumed the to the size of LOCA caus j by a stuck open PORV. steam g and no charging flow makeup to the reactor ccolant systens is in progress.
Three cases were analyzed:
Case l_ [with auxiliary Ieed system (AFS) flow to steam generator and with safety injectica (SI)]
The reactor core remains flooded with cooling water throughout the duration of the analysis (Approx. 4000 sec.)
Results:
a.
and thej arameters indicate that no uncovering of the core p
would occur thereafter.
The prhssurizer steam-water mixture level increases and b.
stabilizes at about a 2/3.
Case 2_ [with no AFS and with SI]
Same'as Case 1, a. and b.
Results:
a.
2/Yof steam generator level is still present at 4000 sec.
b.
Reactor coolant system pressure approaches 1100 psi which I
, -WSorresponds te the temperature in the steam generators c.
with safety valves lifting.
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E. G. Case APRIL 1
1979 Case 3 [with AFS and no SI (i.e., no water makeup to the reactor coolant system at all)]
Results:
a.
The reactor core would start to uncover at about 2100 sec.
Additional analysis is being done by W for Case 3 without the steam generators isolated.
And a comparisoE of Cases 1 and 2 indicate that the results are not very sensitive to AFS initiation for the time periods of the analysis.
E discussed the signals which initiate SI., Analyses which they submitted previously (Zion Station and RESAR-3 dockets) show that a small LOCA in the pressurizer steam space may not result in SI initiation because the pressurizer level may not decrease. A coincident pressurizer low level (Lp).nd low pressure (Pp) is neec'ed for SI actuation. But their analysis of containment building pressure following this LOCA shows that SI would be initiated by containment pressure high (no.1) indication setpoint which is set at abc,.10 a of containment design pressure at about 1600 sec. At 1600 seconds, reactor core fual surface temperature would be at the same temperature as the reactor coolant system coolant which is saturation tempera,Jre for 1100 psi. This is far below the temperature necessary for core damage. To provide additional assurance that SI initiates and prevr~ ts the core from becoming uncovered, in addition to considering the high containment pressure setpoint SI actuation signal, W has instructed its customers that SI snould be manually initiated if Pp decreases to the low Pp setpoint regardless of Lp readina.
j E is still evaluating the question of when to manually shut off SI following its acti"3 tion. The concerns are (?) that the SI system would fill the reactor coolant system complete i 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 of a TMI-2 type incident (core uncovery). W presented a logic " tree" that an operator could use to detemine if SI should or should not be shut off following events which lead to SI and low or failing 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 overormsure as discussed above should be included.
W noted that the bulle ein provision regarding containment isolation reset is not applicable to its plants because containment isolation valves do not open following an SI reset (as occurred at TMI-2) unless the operator deliberCaly opens the isolation valves.
228 232
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.. APRIL 1 2 1973
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E. G. Case Following the W presentation, the staff discussed the followup action A bulletin will probably to be taken in light of the W information.
The bulletin will be i
be issued to W facilities in the next few days.
essentially tiie same as 0I&E Bulletin 79-05A but additional information will be included to de'. ermine plant specific corrective measures dealing with:
- I 1.
Manual shutoff of SI,
,i-Management checking of safety system operability status, 2.
possible elimination of Lp as an SI initiation signal by placing it t
f 3.
I in a " tripped" state, n
Possible requirement for containment isolation on high radiation 4.
k 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.
AAN-U4N)'
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Denwood F. Rcss, Depuky Director '
r Division of Project Management
Enclosures:
As stated cc w/ encl:
See next page t
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E. G. Case april 1 2 579 Distribution f
Docket (50-320)
00R Reading
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NRR Reading
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H. R. Denton V. Stello R. Vollmer 5.:
W. Russell B. Grimes I.
T. J. Ca rter
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I D. G. Eisenhut I-A. Schwencer
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D. L. Ziemann 4
j; P. Check j
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G..C. Lainas f
D. K. Davis h'
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T. A. Ippolito R. W. Reid
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V. Noonan e
G. Knighton J'
l M. Fletcher
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D. Brinkman
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Attorney, OELD l-R. Fraley, ACRS(16)
J. R. Buchanan
'7 TERA NRC Participants j
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Enclosure LIST OF ATTENDEES WESTINGHOUSE MEETING 04/11/79 Carolina Power & Licht NRC D. B. Waters R. S. Boyd, DPM T. A. Ippolito, 00R J. J. Sheppard M. H. Fletcher, 00R Shaw, Pittman, Potts & Trowbridog R. Lobel, DOR E. G. Case, NPR J. H. O'Neill F. Orr, DSS A. Ignatonis, DSS American Electric Powe_r St.
Coro.
N. C. Moseley, I&E J. L. Crews, I&E Region V E. A. Reeves, DOR J. G. DelPeriro N. Anderson, DOR Public Service Electric & Gas E. Wenzinger, 00R M. Mendonca, 00R L. B. Marsh, DDR P. A. Moeller D. Neighbors, 00R Westinchouse_
J. Wetmore, DDR T. V. Wambach, 00R R. W. Stutter A. Burger, DDR B. C. Buckley, DPM K. R. Jordar A. J. Szukiewicz, DSS V. J. Espusito W. J. Johnson J. Guibert, OCM A. Schwencer. DDR T. M. Anderson G. Zwetzig, D0R S. H. Hanauer, DSS Souihern Californi: Edison
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F. Schroeder, DSS L. P. Croker. DPM J. Rainsberry D. '/assallo, DPM A. Thadani, DSS G. Lainas, DDR D. F. Ross, DPM D. G. Eisenhut, DDR 2
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