ML20053A642
| ML20053A642 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/30/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1969, NUDOCS 8205270056 | |
| Download: ML20053A642 (7) | |
Text
/96 9 DATE ISSUED:
3/30/82 6"
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MEETING MINUTES
$NO MARCH 24,1982
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ALBUQUERQUE, NM PURPOSE The purpose of the meeting was to discuss the NRC's use of LOCA/ECCS codes, the impact of the recent Ginna steam generator tube rupture (SGTR) event on LOCA/ECCS-related matters, and NRR's evaluation of operator guidelines and procedures for transients / accidents.
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ACRS NRC s
M. Plesset, Chairman B.Sheron, NRR p]
4y J. Ebersole, Member W. Hodges, NRR s$$%w7 A
C. Mark, Member J. Laaksonen, NRR E
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W. Mathis, Member W. Lyon, RES g
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D. Ward, Member L. Shotkin, RES A. Acosta, Consultant I. Catton, Consultant EG&G V. Schrock, Consultant D. Hall T. Theofanous, Consultant Z. Zudans, Consultant P. Boehnert, Staff *
- Designated Federal Employee MEETING HIGHLIGHTS, AGREEMENTS, AND REQUESTS 1.
Dr. B. Sheron (NRR) discussed the status of the reactor coolant pump (RCP) trip issue.
NRR believes that allowing manual RCP trip will not significantly impact the risk of core melt as a result of a SB LOCA provided certain guidelines can be met and acceptable RCP trip criteria are established.
The Staff is proposing to send a letter to the PWR licensees that wili state the NRC acceptance criteria for manual trip and require licensee response on how they will meet the criteria (Figures 1-2).
Other NRC guidance noted included: trip RCPs only for a LOCA, consider partial or staggered trip schemes (possibly leave 1 or 2 pumps running), and upgrade pressurizer auxiliary spray capability as necessary.
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ECCS Meeting March 24, 1982 5
Dr._Plesset sugcysted that the Subcommittee meet again with NRC on this issue when the Staff has received and evaluated the licensee responses.
2.
Mr. W. Hodges (NRR) provided information on the potential problem of errors in BWR level detectors due to flashing in the instrument lines.
There are two level devices in use - the Yarway (Figure 3), and the newer cold reference leg instrument (Figure 4).
The Yarway is more subject to flashing because the reference leg is located inside containment.
NRC noted that the newer instrument is less prone to flashing and for most of them even, if flashing occurs, the errer is minimized due to small changes in elevation in the drywell. GE is aware of the potential flashing problems and has accounted for it in their proposed operator guidelines.
NRC has issued a preliminary report on this issue for comment by the Staf f, GE, and the GE Owners Group. NRC will review the comments when received and decide what actions need to be taken.
In response to questions from Mr. Ebersole on the survivability of level sensing instruments when subjected to blast effects given a LOCA or steam line rupture, Mr. Hodges said he will provide written information on the survivability of these instruments.
3.
Mr. J. Laaksonen detailed the Ginna steam generator tube rupture (SGTR) event and the plant system response.
Mr. Sheron noted that the informa-tion presented is considered preliminary and NRC is not drawing any con-clusions regarding operator actions at this time.
(An of ficial NRC report l
on the incident is due to be issued soon).
The SGTR event sequence was detailed. Highlights /significant events noted l
included:
- %e operators had positively diagnosed the tube rupture at about 7 minutes into the transient. Se steam jet air ejector high radiaton alarm was a key indicator of a tube rupture.
ECCS Meeting March 24, 1982
- 2e RCP's were manually tripped 50 seconds af ter trip criteria were det in accordance with specific Ginna procedures.
- The PORV was opened repeatedly to depressurize the RCS, again in accordance with procedure, however the valve stuck open on the fourth opening. The block valve was closed to stop the leak.
- R e operators continued HPI longer than necessary. NRC speculated that this resulted from concern over upper head flashing and pressurizer level being off-scale high which were not addressed in the plant procedures.
Failure to secure HPI led to overfilling of the failed SG and relieving of water /two-phase liquid through the steam line safeties.
NRC noted the following regarding the event:
(1) there was extensive 3
upper head voiding during PORV operation (volume extent about 300 ft ),
(2) there was no significant vessel cooling, thus no pressurized thermal shock seen; (3) pressure control using the PORV appears to be difficult; and (4) RCS subcooling was maintained during the time the pressurizer was enpty (or off-scale low) except for possible upper head saturation for a few minutes after reactor trip.
In response to a question from Mr. Ward, Dr. Sheron said only a few plants have automatic RCP trip.
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The nanagement of SGTR events in CE plants without PORVs was reviewed by Mr. Laaksonen. The following key points were noted:
off head.
For low-head HPSI, RCP trip appears inevitable thus conplicating the transient. All CE plants (except Maine Yankee) have low head HPSI.
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- Normal pressurizer spray is probably not available during short-term action in low head HPSI plants. CE plants have auxiliary spray, bitt SGTR guidelines do not mention its use.
ECCS Meeting March 24, 1982
- Containment isolation signals vary in CE plants. 'Ihe guidelines issume no automtic isolation during a tube rupture event.
- A step-by-step co@arison of W and CE S3TR operatot guidelines indicates that the CE guidelines are inco@lete; e..g., nc guidance is provided for stopping the tube leak; thus the leak will continue during plant cooldown.
In response to Subcommittee questions, Dr. Sheron said CE is submitting a revised set of guidelines for NRC review. CE has also been asked to provide a co@arison of how a SGTR can be controlled with and without use of a PORV.
5.
Mr. Lou Shotkin (RES) discussed the status of LOCA/ECCS code development for use by NRR. He noted that structure development will end for the advanced LOCA/ECCS codes in FY 83-84; however maintenance of the code correlations will continue. The focus of effort beyond FY 83 will be on development of a plant analyzer and establishment of a co@rehensive DVR data bank.
The question of why NRC uses two adyanced system codes (RELAP-5 and TRAC) with overlapping capabilities was discussed.
RES believes that use of both codes provides a cross-check for potential major errors, a cross fertilization of ideas and models and, since many utilities are using RELAP-5 for licensing, TRAC can be used for audit calculations.
RES, detailed the main capabilities and limitations of the LOCA/ECCS system codes (TRAC IMR and BWR, RELAP-5, COBRA / TRAC) as well as their use in the regulatory process. Figures 5-10 sumnarize the discussion.
Mr. Shotkin requested advice from the Subcommittee on two item:
1.
Should RES incorporate nulti-deminsional kinetics in the TRAC PWR and BWR codes?
ECCS Meeting March 24, 1982 2.
RES wants to establish an LWR data bank in order to develop input decks for use in calculating IMR transients with the plant analyzer.
RES wants advice on how to obtain the necessary plant data from the licensees.
Dr. Plesset requested that the Subcommittee Members and Consultants provide their advice on these two items to the DFE in the near future, for transmittal to RES.
6.
Mr. W. Lyon (RES) discussed a new version of REIAP-5 (ZEIAP) that has been developed for NRR application. ZELAP is a version of RELAP-5 MOD-1 with inprovements in reflood nodeling, boron reactivity modeling, separator nodel, jet pump model, and the subcooled boiling model. The code will be conpleted and verified in mid-April 1982 and will be used until the MOD-2 version is conplete in 1983. ZELAP will also be subjected to independent assessment.
A large-screen video denonstration of the code's capabilities was given. A sinple PWR IDFt-type plant transient was rrn (loss of feedwater was the initiator - Figure 11). 'Ihe code has interactive capability (turn pumps of f/on, open/ close valves, etc.) which was also demonstrated.
Dr. Plesset indicated that this code capability has promise and further development of this capability should be encouraged.
7.
The use of BWR and PWR LOCA/ECCS codes by NRR was discussed by B. Sheron and W. Hodges. Dr. Sheron said that NRR has developed an extensive indepen-dent audit capability and has also begun a program to develop input data decks for most classes of reactors. NRR will be deenphasizing the LOCA code effort in favor of codes used to calculate plant transient response for events including nultiple failures and systen interactions.
8.
Review of operator guidelines and procedures by the Reactor Systens Branch (RSB) was detailed. RSB reviews the guidelines for technical adequacy.
The human factors aspects of the guidelines are reviewed by the Division of Human Factors. The RSB review focuses on two central criteria:
(1) fuel integrity, and (2) RCS pressure.
In response to Subcommittee questions on
ECCS Meeting March 24, 1982 g
the guidelines, Dr. Sheron suggested the Subcomittee may benefit from a discussion of this item with the representative Owners Groups and vendors.
Mr. W. Hodges discussed review of the BWR guidelines. Key aspects of the guidelines include:
- The guidelines are symptom, rather than event., oriented.
- They are generic to BWR/l through BWR/6 plants in that they address all major systems which may be used to respond to an einergency.
- Entry conditions for the guidelines are symptomtic of both emergencies and events which may degrade into emergencies.
- W e guideline format is based on two funcitons, a RPV level control guideline, and a containment control guideline.
Mr. Hodges said experience to date with the new guidelin L has shown them to be an irprovement over the old procedures.
The PWR guidelines and procedures were discussed by Mr. Laaksonen. He noted that among other things the revised guidelines are expected to address events beyond the design basis.
Current status of the PWR guideline development was noted.
In general, the W guidelines are being revised by the W Owners Group to meet the NPC post-TMI requirements. m is program should be complete by mid-1982. The B&W guidelines are being developed on a plant-specific basis. NRC and l
B&W have agreed that the guidelines developed for Oconee Unit 3 should serve as the lead document for the other plants. The Oconee 3 guidelines should be complete in April 1982. CE is scheduled to submit a revised set of guidelines in April 1982 for Staff review following NRC concern that the original guidelines were event, not symptom, oriented.
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ECCS Meeting March 24, 1982 t
Dr. Plesset commented that the effdrt on development of these guidelines appears on a sound footing, an'd he. encouraged continued effort in this area.
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Themeetingwasadiournedat4:05p.m)
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NCTTE: Additional details can be obtained from the transcript located in the Public Document Room; 1717 H Street, N.W., Washington, D.C. 20555 or from Alderson Reporting, Inc.,,400 Virginia Avenue, S.W. Washington, D.C.
(202) 554-2345.
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