ML20141E924
| ML20141E924 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/02/1985 |
| From: | Bender M Advisory Committee on Reactor Safeguards, QUERYTECH ASSOCIATES |
| To: | Shewmon P Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-CT-1822, NUDOCS 8601080425 | |
| Download: ML20141E924 (3) | |
Text
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/j 3/b QUERYTECH A550CifEES Engineering and Technical Advisors 9040 6scttfw Park Driw Suite 217 Hnoxville, Tennessee 37923 Phone: (615) 690-2728 11_..
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Dr. Paul Shewmon Chairman Millstone 1 Subcommittee Advisory Committee on Reactor Safeguards Washington, D.C.
20555
SUBJECT:
Comments on Millstone One Subcommittee Meeting'at the Plant Site, November 18 and 19, 1985
Dear Paul:
The information developed at the November 18-19 Subcommittee Meeting was helpful in establishing how operating ~1icenses should be considered for the older nuclear power stations.
Since Millstone was completed and put into operation in December, 1970, and came to full power on January 3, 1971, it is representative'of a class of nuclear power plants that was built'in the era before the current intensive campaign to establish a complete set of inspection records for plant construction.
Recognizing that the plant construction started in May, 1966, the record shows that a less than five year construction period is adequate to complete one of these installations.
The operating record has been remarkably free of serious operational problems, although, like other boiling water reactors, Millstone 1 has had its share of stress corrosion cracking problems.
Corrective methods have been the same as those generally applied to operating plants.
At Millstone 1, these measures have generally been adequate without the need to replace the primary circuit piping.
In these III units, of which Millstone 1 is~the most successful example, the only really serious difficulty has been associated with cracks in the feedwater spargers, a chronic problem in boiling water reactor systems.
Corrective measures taken at Millstone have apparently been successful.
There has been no evidence of feedwater sparger cracking for the past two years.
The successful operation of the plant certainly justifies the granting of a full-term operating license.
The comments I wish to offer concerning Millstone 1 are directed mainly to the probabilistic risk assessment (PRA) program bein developed at Millstone and the NRC staff's integrated safety assess g E E O C N 'L 1
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ment program (ISAP) being implemented in parallel with the full-term operating license review for Millstone 1.
The ISAP program is a follow-on to the systematic evaluation program (SEP) that has been active for several years, covering representative examples of the older nuclear power plant installations that continue to operate.
ISAP adds to the SEP program certain regulatory requirements that have been imposed by rule making and regulatory orders, including the TMI-2-Lessons-Learned Action Plan.
ISAP is.a major step forward in bringing about a consolidated review of the safety adequacy of the
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older operating units.
A few significant items remain outside the ISAP program, including seismic events, external flooding, tornados, and sabotage.
Millstone is in conformance with the regulatory requirements as presently interpreted for these external events, but the regulatory staff has not attempted to make these matters a part of the ISAP effort.
The ISAP takes advantage of the Northeast Utilities' probabilistic safety study effort that is directed to safety issues of concern i
to both the NRC and the Millstone operating organizations.
The earlier Millstone 3 probabilistic safety study (PSS), a " level 3" PSS equivalent to the most detailed probabilistic risk assessment (PRA) performed by other utilities,.was reviewed by ACRS for the Millstone 3 operating license.
The Millstone 1 effort is less detailed than the Millstone 3 work, but includes nearly all of tha important internal safety issues that should be addressed in such study.
It also takes advantage of the IDCOR work, wherein the sequence of events associated with postulated accidents are carried to their logical conclusion by phenomenalistic analysis.
Probabilistic aspects of the event are generally judged on the basis of operating experience asscciated only with the initiating events in combination with the Wash 1400 event tree probabilities, now generally in use.
While not enough information was available to directly evaluate the statistical validity of these judgments, one could conclude that, i
barring some unintentional oversight, this level 1 effort combined with the Millstone 3 work should give adequate safety insight to make a judgment concerning the relative safety Millstone 1 as compared to i
other BWR systems, now operating, that were constructed in the same time frame.
i Millstone 1, as with other probabilistic studies by Northeast Utilities, has used operational experience at the Millstone site and adjusted its statistical judgments when they were derived from other studies, to take units to the broader statistical base. account of the relevance of the Millstone If such adjustments are made cautiously with due consideration to the limited sample size from a i
2-unit reactor station, the approach should be suitable.
In listening to the discussions by Dr. Bickle about the Millstone 1 PSS, I concluded that one should make certain that the data being used does not unintentionally overlook the experience limitations that exist because some important accident events involve equipment
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basis.
Millstone 1, for example, is' using its own experience for loss-of-off-site power as a basis for judging the statistical frequency of such events.
It conservatively assumes that circum-
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i stances like the recent east coast hurricane, Gloria, will cause loss of off-site power even though the event in question was anticipated and provisions had been made to shut down the reactor prior to experiencing the hurricane effects.
The resulting statistical base in that case is probably representative of typical nuclear plants on the eastern seacoast, but there may be other events where the statistical interpretation is more or less conservative.
Relief valve reliability and feed water supply provisions are the types of matters that warrant scrutiny.
It was noted that Millstone 1 has not t
yet performed an internal flooding analysis.
This is a notable ommission because it is one of the more controversial questions associated with safety adequacy.
The Millstone organization plans to perform such a study.
With regard to the ongoing work for ISAP and PSS, the Millstone effort is bringing about a more logical interpretation of the Appendix R fire protection requirements and environmental qualifica-tion of safety related equipment.
Previously, the interpretations of these requirements have been somewhat arbitrary and occasionally resulted in less-than-satisfactory resolution of the interpretation controversies.
I believe this effort should be encouraged and that it could be expanded to address many of the seismic design issues that are arbitrarily treated.
It could also be used to clarify many of the construction quality questions that remain outside the several PRA programs.
The ACRS should encourage more probabilistic treatment of the broaden, construction quality and seismic design questions in order to the perspective associated with their safety significance.
As a final comment, it is worth noting that the operating experience of Millstone is as good and perhaps better than that of nuclear power plants of a more recent vintage where there has been much more intensive emphasis on inspection practices and construction quality.
Although the older plants have not really experienced the accident conditions where many of these quality issues could be of i
significance, there does not appear to be any logical reason for expressing concern about the newer plants if the older ones are operating safely and the regulatory authorities do not see a reason to be concerned for their safety adequacy.
Sincerely, M. Bender MB/kir cc:
John Shiftkins ACRS Staff for ACRS Distribution 3
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