ML20043D685
| ML20043D685 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/08/1990 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2689, NUDOCS 9006110068 | |
| Download: ML20043D685 (9) | |
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$52 cu Summary / Minutes Subcommittee on the Systematic Assessment of Experience Proposed Power Level Increase for Indian Point Nuclear Generating Station, Unit 2 l
February 6, 1990 Actions, agreements and commitments, subcommittee on the systematic Assessment of Experience, February 6, 1990.
1.
The subcommittee agreed to bring the proposed power level increase for Indian Point Unit 2 to the full committee.
2.
The subcommittee agreed to propose a letter on the power level
-increase for Indian Point Unit 2.
The ARCS subcommittee on the Systematic Assessment of Experience met on February 6, 1990, Room P110, 7920 Norfolk Avenue, Bethesda, Maryland at 8:30 a.m.
The subcommittee met to discuss the review of the proposed power level increase for Indian Point Nuclear Generating Station Unit 2.
Dr. H. W. Lewis was the subcommittee's chairman for this meeting. The other ACRS members in attendance were:
Mr. Carlyle Michelson, Mr. Charles J. Wylie, Mr. David A.
Ward, Dr. Ivan Catton and Mr. James C. Carroll. Herman Alderman was the cognizant ACRS staff member for this meeting.
Dr. Lewis noted the discussion would be about the proposed power level increase for Indian Point 2, which is an appropriate 11 percent increase.
The format to be followed would be to hear from the NRC staff, then the licensee, and then the subcommittee would decide what action to take.
Dr.
Donald Brinkman noted that Mr. Cowgill, Region I, will discuss operational experience at Indian Point 2.
This will be followed by the licensee's LE ~~.iATED ORIGINAL 9006110068 900308 o
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c Minutes / Systematic Assetsment 2
of Experience - Feb. 6, 1990 presentation. Then the staff will present an overview of the evaluation.
Mr.
Tim Col,line will discuss the ECCS evaluation.
Mr. Robert Hermann will discuss the steam generators.
Mr. Brinkman will present concluding remarks.
Mr. Curt Cowgill, project section chief, Region 1.
Mr. Cowgill noted the steam generator dryout of January 1988. This occurred following a refueling outage during the reactor start up.
During the plant heat up, a high rate of steam leakage through the steam generator and
,its associated MSIV, with the lack of normal makeup capability in the steam generator resulted in a total loss of inventory, a dryout, in the steam generator. The dryout took about 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and existed for an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being recognized by the utility management.
When plant management became aware of the event, appropriate analyses were conducted to assure that recovery actions were taken to refill the steam generator such that it would not result in equipment damage.
In response to a question regarding low level indications or alarms, Mr.
Peter Kelley replied that there were n' arms. He noted at the time, the plant was being started up and there were a number of alarms going off.
Mr. Kelley said the utility was concerned about adding cold water to the steam generator.
This was resolved by adding warm water from one of the the other steam generators.
Mr. cowgil.1 noted that Indian Point had a high trip rate in the middle 1980's.
He noted that the SALP reviews from 1986 to 1990 have shown a drastic reduction in the trip rate. He noted that operator professionalism has been greatly improved. He said the region has seen a trend of improved performance
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Minutes / Systematic Assessment 3
of Experience - Feb. 6, 1990 with respect to procedural compliance.
Mr. Cowgill noted improvements since the steam generator dry-out.
He noted the shift turnover proceos has improved. He noted the improved operator professionalism. He noted improved communication within the shift and with plant management.
Mr. Kelley added that Consolidated Edison is in an increasing or improving trend with regard to operations.
In conclusion, Mr. Cowgill noted that Indian Point II operating experience and performance from 1988 to 1990 is adequate to support safe power operation at the higher power level requested by the licensee management.
Mr. Charles Jackson, manager, Nuclear Safety and Licensing.
Mr. Jackson pointed out that one of the original concerne for operation of large reactore and why Indian Point 2 was originally licensed at the 2758 level was to permit accumulation of experience.
Mr. Jackson mentioned that although the original application was for 2758 megawatts thermal, all of the original evaluation for the engineered safety features were done at the 3216 level.
Mr. Jackson pointed out that over the past several years, they have been reanalyzing for reloads, and they have been redoing the safety analyses and using more up-to-date techniques.
Kr. Jackson discussed the procedural changes and training scope. He noted that virtually all of the plant procedures have been reviewed and changes identified for operating procedures. The operators will be trained, starting in about a week, on the changes in the various set points and new
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s Minutes / Systematic Assessment 4
.cf Experience - Feb. 6, 1990 procedures. The simulators are being programmed for the changes..
Mr.
Jackson discussed bioinuling.
He noted that periodically the intake screens are clogged with river grass.
He noted that this requires more frequent cleaning of the intake structures. He noted a minor problem with barnacle fouling. This requires more frequent screen cleaning. He mentionJd a problem with microorganisms that were believed to cause pitting.
The corrective action was chlorination of the service water system.
Mr. Lou L..Beratori, manager of Safety AsJessment, Consolidated Edison, discussed the results of the stretch power program.
He noted that they have performed the large break LOCA and small break accidents and the peak clad temperatures for both are below the 2200 degree i
acceptance of 10 CFR 50.46.
He noted that containment was reanalyzed at LOCA condition for the higher power level. The current value is 41.1 psi versus the previous value of 40.6 psi.
This is well below the containment design pressure of 47 psi.
He noted the staff's origir.a1 SER specified that the containment design pressure should exceed the peak pressure by at least 10 percent. This is clearly the case.
Mr. Michelson asked the staff if you raise the power at a reactor 10-12 percent, are you required to do a natural circulation test or can you depend on the results of other plants.
Mr. Tim Collins, Reactor Systems Branch replied that a plant increasing its power by 10 percent would not be required to do a natural circulation They can depend upon the results of other plants if they like.
test.
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of Experience - Feb. 6, 1990 i
Mr. Jackson noted that they are relying on the Diablo Canyon tests which were performed at a much higher power level than Indian Point is requesting.
He noted that Diablo Canyon was a similar design and a similar configuration.
Several of the committee members questioned whether the plant requirements that have been imposed since the Three Mile Island accident have been reviewed and ensured that the plant power increase meets these new requirements?
Mr. Jackson assured the subcommittee that Consolidated Edison has identified all the additional requirements and incorporated the various modifications in the assumptions for the rsanalysis work that has been performed for this application.
Mr. Liberatori listed the non-loss of coolant accidents that have been reanalyzed:
- Containment performance - Steamline break has been reconfirmed on the limiting case' offsite dose evaluation - has been reevaluated and found to be within the guidelines' Tube rupture accident - has been reanalyzed and found to be well within the guidelines.'
Fuel handling accident - has been reevaluated and found to be within the guidelines' Feedwater temperature - has been increased about 15 degrees and now is about 430*F In summary the limiting FSAR events are now consistently analysed at the
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of Experience - Feb. 6, 1990 3216 power level source term and the results remain below the Part 100 guide-line limits.-
Mr. Liberatori noted that they had looked at the balance of plant systems and reviewed the equipmens against the stretch conditions and determined that the design envelopes the anticipated operating conditions at stretch.
Mr.
Liberatori concluded l.
- The stretch program has demonstrated compliance of the FSAR analyses f
with applicable acceptance criteria' 1
They have demonstrated' compliance of the components and systems with FSAR functions and regulatory _ requirements
- l The stretch program has reconfirmed the capability of the plant to perform in its original guaranteed power rating' They have determined that there is no significant hazards consideration involved as per 10 CFR 50.92C l
Mr. Bram discussed the steam generator dryout la his concluding remarks.
He noted that the operators recognized that the steam generator was drying out. He noted that the operators were aware that the motor driven auxiliary feed water pump was out of service for maintenance.
The operators were given a schedule that indicated that the auxiliary feed pump would be returned to l
service very shortly, and they fully expected the auxiliary feed water pump to be returned to service before the steam generator dried out.
The corrective actions, as a result of this event include additional senior management involvement in plant operation and policies set by senior management were communicated more effectively down to the operating level.
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.Y Hinutes/ Systematic Assessment 7
of Experience - Feb. 6, 1990 The shift technical advisor is now required to be in the control room around the clock to provide additional technical guidance to the watch. The shift watch
- supervisor has been instructed to make sure that he or she doesn't get distracted by small details and to maintain an oversight-of what is going on.
An additional staff group has been established to perform pre-operational L
planning. This group reviews every evolution of the plant and makes a deter-mination whether or not the plant is in a condition to initiate that evolution.
Mr. Wylie asked if after the steam generator dryout they.had noticed any differences in the deterioration of the steam generator that had dryed out L
I compared to the others?
L Mr. Brown replied that they had not.
L Mr. Brinkman, NRC, noted that the staff had concluded in its safety evaluation report that the plant was in fact designed for the core power level of 3071.4 megawatts thermal.
He noted that there wasn't any technical reason for the original power level of 2758 mwt.
The reason was to gain experience at higher power levels than the previous plants.
l The current safety evaluation looked at the core design and determined that it was adequate to perform at the 3071.4 mwt level. The licensee l
evaluated all the FSAR Chapter 14 events and the staff confirmed that the results are acceptable.
Mr. carroll postulated a failure of the primary and secondary eldes of the steam generator and asked if the staff had evaluated that?
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'l Minutes / Systematic Assessment 8
of Experience - Feb. 6, 1990 The staff responded that they had not.
i Mr. Tim Collins, NRC, discussed the staff review of the ECCS system. He noted the staff had done a system overview to see if there was anything particularly different about it, relative to other plants. He noted that the staff looked at the original design rating of the pumps in the system to e-a if the licensee was trying to squeeze more out of them than they were originally designed for.
The staff did a performance analysis of the ECCS system. He noted the results satisfied Appendix K and 10 CFR 50.46 and the staff.found that the ECCS is acceptable for operation at the 3071.4 megawatt thermal level.
l Mr. Robert Hermann, NRC, Materials Engineering, discussed girth weld cracking at Indian Point 2.
He noted the problem started in 1982 with a leak through the steam generator shell.
He noted that there was an area of discontinuity in the shell and cold water impinging on the shell. He noted that repairs of cracks were done in 1987. About a year later, the licensee found severe cracks in the steam generator.
Mr. Hermann said the mechanism being postulated for the cracks in' corrosion assisted fatigue. He said that thermal cycling and contaminants like oxygen and copper contribute to the cracking. He noted the down cover location was thought to aggrevate the situation.
Corrective actions have included removal of the down cover, crack grinding, and weld repair of the deeper areas, and heat treatment of the weld repaired areas. Water chemistry has. been improved, flow conditions have been changed to minLmize cold water flow against the shell, and copper sources have
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of Experience - Feb. 6, 1990 been removed from the system.
He noted that ultra-sonic testing has been improved so that cracking can be detected from the outside of the steam generators.
Dr. Lewis asked the subcommittee members to submit their views on the Y
power increase to him on Thursday morning and he would draft a possible letter. He said the committee would discuss a possible letter on Friday. He notec he would present a summary to the Conunittee on Friday.
The meeting adjourned at 12:12 p.m.
NOTE:
A transcript of the meeting is available at the NRC Public Document Room, Gelman Bldg. 2120 "L" St., NW., Washington, D.C. Telephone (202) 634-3383 or can be purchased from Ann Riley & Associates, LTD., 1612 K St.,
NW.,
Suite 300, Washington.
0.C.
20006, (202) 293-3950.
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