ML19246A362
| ML19246A362 | |
| 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 7906180627 | |
| Download: ML19246A362 (5) | |
Text
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FJUCLEAn REGULATORY COMMISSION 4
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E. G. Case. Deputy Director y'g.
MEMORANDUM FOR:
Of fice of fluclear Reactor Regulation D. F. Ross, Deputy Director FROM:
Division of Project Management
SUMMARY
OF MEETING WITH CC".BUSTION ENG CORRECTIVE ACTIONS FOR COMBUSTION ENGINE
SUBJECT:
PLANTS AS A RESULT OF THREE MILE ISLAND UNIT 2 INCIDENT the NRC staff met with respresentatives of On April 11 and 12,1979, Combustion Engineering, Incorporated (CE) in Beth water reactors (PWR) as a result of the incident at Three A list of Several CE FWR licensees were in attendance.
Unit 2.
attendees is attached (Encicsure 1).
Acril 11, 1979 Meetine The meeting opened with an overview of the events at Three Mile Islan Unit 2 (TMI-2) which require imediate attention by all operating PWRs as these events are perceived by the staff in light of informai available at this time.12 in the NRC Office of Inspection and Enforcem 79-05A of April 5,1979 (Enclosure 2).that the responsibility f these items rests with CE and the utilities.
that are needed are specific instructions to be issued imediately to These corrective measures will be reviewed by D. Ross of licensees of CE PWRs.
the NRC staff and issued by means of an 01&E Bulletin.
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the NRC staff read through the items in the B&W bulletin (Enclosure and asked for coments and agreement that the items were suitable for CE designed plants.
Items 1 through 3 were agreed to by CE (agreement meaning that the item was appropriate for a CE designed plant).
They concern the overriding of automatic Item 4 consists of f)ur parts.
actions of engineered safety features (ESF).
CE was unwilling to speak on this item because they consid prerogative of their customers. lines for the development of procedure
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CE stated that SI actuation occurs on the following signals (among others):
1.
low primary pressurizer pressure 2.
high containment pressure Containment isolation is actuated by the high containment pressure (with the same set point as for SI actuation).
CE ' stated that while pressurizer level is the main parameter looked at in the guidelines, other system parameters are also used by some CE customers.
g CE has not provided a guideline for turning off the safety injection J
to its customers.
CE agreed with item 4a (the' operator should not override automatic actions of emergency safec..rds features).
A representative of Baltimire Gas & Electric Company questioned the reason for a time require..ent in item 45 concerning conditions under which the HPI could be turned off. He cJestioned why a prcisure i
I indication was not suffic ent. The point was that if the neactor Coolant System (RCS) were to approach going solid, pressure (subcooling) indicaticn would be sufficient, regardless of the time that the HPI had been in operation. The staff noted his co m.ent.
The ouestien of the HPI causing reactor vessel pressure in excess of the allowable limits was discussed. CE stated that they had performed some fracture mechanics calculations of this type for the steam line break for vessels of different ages and these showed acceptable results.
The analyses were previously reported to the staff in the letters listed below.
g 1.
June 4,1975 letter from W. Corcoran, CE to R. Maccary, NRC.
P.
June 24, 1975 letter frcm W. Corcoran, CE to F., Schroeder, NRC.
CE stated that reactors designed by them have not experienced a stuck open relief valve.
For this event, the pressurizer level could increase, -
although the primary system would be depressurizing.
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.. APRIL 1 2 1973 E. G. Case CE related a case in which the system drained down to a pressurizer level of 1% due to a drain valve that was inadvertently left open during There was no flashing or change of direction in a test (pre-nuclear).However, this event was not directly applicable pressurizer level.to the discussion since the drain was releasing water inventor the water space r;ther than steam space.
CE stated tha+ the analysis of the inadvertent opening of a relief valve showed that the two phase level reached the top of the pressurizer with a void fraction in the pressurizer of approxiamtely 25%.
CE stated that under this condition they would expect the pressurizer level instrumentation to give a true indication of level.
I The staff asked whether the operator had to take any action based on CE responded that he did not and furthermore, that level did l evel.
not enter into any safety system actions.
The staff then asked CE if.'the pressurizer could be full while the CE replie. that the analysis of the inadvertent core was empty.
opening of a relief valve snowed that the pressurizer would drain.
The staff questioned whet' er the computer code used for the analysis CE stated that the code was would have predicted this phenomenon.The staff will pursue this further with CE.
an evaluation model code.
a This completed the discussion of Item 4.
Item 5 concerned verifying that the auxiliary feedwater (AFW) valves CE agreed with this item and stated further are in the proper position.
that the AFU systems on all CE operating plants are manually actuated.
Analyses of all transients in the FSAR showed that wit I
not open.
g CE agreed that items Fythrough 10 of the OI&E Bulletin were appropriate.
CE had stated earlier that they had rot come prepared to answer these They stated that the answers they gave could be items in detail.
considered those of the Corporation however.
The staff held a caucus and decided that more information was required in order to write the bulletin, and CE was given an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to make comments in the following areas:
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. APRIL 1 2 079 E. G. Case 1.
ESF reset criteria What temperature What criteria are now used for turning off HPI?
2.
and pressure criteria are used?
Should NRC advise utilities on CVCS operation?
3.
4.
Further discussion of fracture mechanics.
5.
Criteria for tripping the reactor coolant pump.
List operator actions based on pressurizer level.
6.
I Provide a CEFLASH calculation for a small treak on the pressurizer 7.
fluid dynamics.
g It was decided to continue the meeting on April 12,1979, at 1:30 pm.
April 12,1979 Meeting
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cview of the NRC intentions to isuue bulletins The meetino opened with a to Westinghouse Electric Corporation and Combustion Engineering reactor licensees by tctorrow. Ap-il 13,_1979.
It was noted that Bulletin 79-06 was issued on Apfil 11, is79, and contained general guidance for all PWR reactor licensees (except Babcock & Wilcox plant licensees).
The bulletins to be issued temorrow v.'ill be based on but provide more specific guidance than Bulletin 79-06.
l The meeting then proceeded to the seven point agenda identified at the end of the previous day's meeting. These seven points correspond to the provisions of item 4 of IE Bulletin 79-05A.
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The CE representatives discussed these provisions as follows:
b 4.a. (agenda item 1) CE agreed that this was accropriate for their facilities and suggested additional clarifying instructions to The staff will consider this suggestion.
the plant operator.
4.b. (1) and (2) (agenda itecs 2, 3 and 4) CE stated that the they confirmed the information given to us in the April 11 meeting
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Based on their investigations and regardino fracture mechanics.
11, 1979 theM' ~information (identified in the April mirutes), they agreed that the provisions of item 4 were appropriate for their facilities.
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ifRIL 1 2 1373 E. G. Case.
4.c. (agenda item 5) The CE representatives believe that additional clarification should be added to this provision such that not all reactor coolant pumps should be run, but at least one pump per cooling loop should be run.
It was recEgnizeo da, ici A...e facilitias, the creviso that cr.a pump por ! con ha run weald raquirn all reactor coolant pumps to be run. The staff will consider this addition.
4.d. (agenda item 6 & 7) CE agrees wi;.1 this provision as it is presently worded. Regarding their emergency core cooling code (CE FLASH) calculation for a small pressurizer break, CE representatives state that some information is available on the CE System 80 dockec (Safety Analysis Report Chapter 6).* This information provices pressurizer pressure as a function of time.
CE will, after a check of proprietary considerations, make available informa tion regarding pressurizer level as a function of time for this small pressurizer steam space break (equivalent to a 4" dia. hole).
Based on the information presented by CE, the staff intends to proceed with issuance of a bulletin with short term corrective actions to Cr facility licensees.
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Denwood F. Ross, Ceputy Director Division of Project Management t
Enclosures:
As stated cc w/ encl:
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