United States Nuclear Regulatory Commission - Protecting People and the Environment

Request for Additional Information Assessment of Licensee Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells (Generic Letter No. 87-05)

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D. C. 20666





In 1980 at the Oyster Creek Nuclear Generating Station GPU Nuclear
Corporation  (GPU) observed water coming from the drains that are
connected to the sand  cushion located between the drywell wall and the
surrounding concrete. The  licensee continued their investigations to
locate and correct the source of  leakage during the 1983 refueling
outage. In the spring of 1986 during the  1986 refueling outage the
licensee again noticed water coming from the drains.  At that time they
decided to perform ultrasonic thickness measurements of the  drywell shell
plates adjacent to the sand cushion to determine if corrosion  had
occurred. On November 20, 1986, the licensee informed the NRC of their 
initial ultrasonic examination findings. The staff issued Information
Notice  No. 86-99 on December 8, 1986 providing details on the source of
leakage,  actions taken by the licensee to eliminate the leakage and
preliminary  information on the extent of corrosion that had occurred on
the drywell shell  in the sand cushion location. From the results of
further non-destructive  examinations of the drywell shell, metallurgical
analyses of samples removed  from the drywell shell and chemical analyses
of the sand and water obtained  from the sand cushion region, the wall
thinning appears to have resulted from  general wastage of the carbon
steel plates from corrosion caused by water  containing aggressive anions.
From the UT results it appears that general  wastage of the shell, to a
varying extent, has occurred in two localized  regions of the drywell
shell plates adjacent to the sand cushion. One  localized region extends
about 35 feet circumferentially, the other region  extends about 21 feet
circumferentially. The drywell shell in these regions  was reduced from a
thickness of 1.115 inches to an average of about 0.850  inches with some
local spots being reduced to about 0.750 inches. The staff  issued a
Safety Evaluation on this subject by letter to GPU dated December 29, 
1986. The staff, through contacts with the BWR Owner's Group and various 
licensees has completed a survey to obtain additional information on the
sand  cushion details, the gap materials and the inspections performed at
various  facilities. A summary of the results are presented in Table 1.


The purpose of this Generic Letter is to obtain information from BWR
Owners  regarding their intended actions to determine if drywells at their
facilities  have degraded by the corrosion mechanism described above and
if the licensees  have current and/or future plans to minimize the
potential for this problem at  their facility. The information provided
will be utilized by the staff to   determine, pursuant to
10CFR50.55a(9)(6)(ii), if an augmented inservice  inspection program is
required for the Mark I drywell shell to assure its  continued structural


From the information obtained from the review of the Oyster Creek event,
it  appears one source of leakage that resulted in wetting the sand
cushion was  related to the deterioration of a drain line gasket at the
drywell to cavity  seal (see figure 1). There also is some question as to
whether or not the  drain line which serves to remove any leakage past the
drywell to cavity seal  was functioning properly.

Upon reviewing the details of the sand cushion design at the various Mark
I  facilities, there appears to be a design feature that varies, dependent
on the  architect engineer, that may be significant. In some designs,
Figure 2, the  sand cushion is covered with a galvanized steel plate which
is sealed to the  drywell shell and the surrounding concrete. Further,
drains are provided to  remove water which might collect on the plate from
above. In other designs,  Figure 3, the sand cushion is open to the gap
between the drywell shell and  the surrounding concrete and the only path
for water to drain from the sand  cushion is through the small drain lines
at the bottom of the sand cushion. It  appears from the Oyster Creek
experience that, with the high permeability of  the sand cushion and the
small bottom drains, the likelihood of adequately  drying the sand cushion
is low if copious amounts of water have entered the  sand cushion. The
staff has sufficient concern to initiate the collection of  information of
the licensee's current or proposed action to assure the  degradation of
the drywell shell plates adjacent to the sand cushion has not  occurred
and to determine if augmented inspections above and beyond those  planned
by the licensee's are necessary.

Information Requested

Pursuant to the provisions of 10CFR50.54(f), licensees are requested to 
provide the following information under oath or affirmation to the Office
of  Nuclear Reactor Regulation:

All Mark I Owners

1) Provide a discussion of your current program and any future plans for 
determining if the drain lines that were provided at your facility for 
removing any leakage that may result from refueling or from spillage of
water  into the gap between the drywell and the surrounding concrete or
from the sand  cushion itself are unplugged and functioning as designed.


2) Provide a discussion of preventive maintenance and inspection
activities  that are currently performed or are planned to minimize the
possibility of  leakage from the refueling cavity past the various seals
and gaskets that  might be present.

3) Confirm the information listed in Table 1 is correct with regard to
your  facility.

Mark I Owners Whose Designs Are Such That The Sand Cushion Is Open To Gap
Between The Drywell Shell And Surrounding Concrete

Provide any plans for performing ultrasonic thickness measurements of the 
drywell shell plates adjacent to the sand cushion or any other proposed 
actions to ascertain if plate degradation has occurred. Since the
degradation  that has occurred at Oyster Creek is localized, sufficient
details should be  included to show that the sampling basis for ultrasonic
thickness measurements  is adequate in terms of size and test location.

Licensees and applicants are requested to respond to this generic letter 
within 60 days of the date of this letter. Our review of your submittal of 
information is not subject to fees under the provision of 10CFR170.

This request for information was approved by the Office of Management and 
Budget under clearance number 3150-0011 which expires September 30, 1989 
Comments on burden and duplication may be directed to the Office of
Management  and Budget, Reports Management Room 3208, New Executive Office
Building  Washington, D.C. 20503.


Robert M. Bernero, Director
Division of BWR Licensing
Office of Nuclear Reactor Regulation

As stated

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