ML20246E632
| ML20246E632 | |
| Person / Time | |
|---|---|
| Site: | University of Buffalo |
| Issue date: | 05/03/1989 |
| From: | Henry L BUFFALO MATERIALS RESEARCH CENTER |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8905110320 | |
| Download: ML20246E632 (9) | |
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'o- o BUFFALO MATERIALS RESEARCH CENTER 2.) L.%
May 3, 1989
.U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 :
RE: License R-77 l Docket 50-57 Gentlemen: ;
Enclosed is a' report regarding the " reportable occurrence" at the Buffalo Materials Research Center, of the State University of New York at Buffalo, on 4/23/89. This report augments telephonic com-munications with NRC staff at Region I on 4/24/89.
Questions regarding this report should be directed to me at '
716-831-2826.
Yours very truly, l Louis G. Henry, Jr. <r !
Director LGH:glvb enc.
cc: Ted Michaels Falter Baunack I
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8905110320 890503 PDR ADOCK 05000057 1 S PDC \
SUNY-BUFFALO, ROTARY ROAD, BUFFALO, NEW YORK 14214-TELEPHONE (716) 8312826
- Easvlink 62910144
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REPORTABLE OCCURRENCE AT l BUFFALO MATERIALS RESEARCH CENTER LICENSE R-77 DOCKET 50-57 i
Introductieg I On 4/23/89 the Operations staff of the Buffalo Materials Research Center (BMRC) completed the required annual measurement of the containment building leak rate. Technical Specification requires that the leak rate of the containment not exceed 7 cubic feet per minute (CFM) of standard air, at a negative pressure differential of .5 inches of water (.018 PSIA). The leak rate is measured annually per. Technical Specification 4.4.(3), and in accordance with Operating Procedure 81 . The measurement is accomplished by drawing tha building to a negative .5 inch pressure and measuring the volume of air which must be drawn from containment to maintain the negative pressure over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
The leak rate established by the test which concluded on 4/23/89 was approximately 19 CFM at .5 inches. The Operations Manager reported the test results to the Director immediately. Since the leak rate exceeded the Technical Specification limit operation of the reactor was suspended.
On 4/24/89 testing was performed to determine the cause of the excessive leak rate. Two small leaks were identified from the personnel airlock gaskets (leakage of compressed air into containment). A larger leak was discovered in the 36 inch containment exhaust isolation damper. It is believed that this leak explains most of the 19 CFM leak rate.
In the course of searching for possible sources of leakage a design error was discovered in the air exhausted by this duct. A monitor air by-pass which is used to regulato flow through the monitor could potentially increase the rate of release of airborne radioactivity in the event of an accident requiring containment isolation. This design error is easily correctable as described later in this report.
DApis for Technical Specification Limit i
The requirement to maintain the building leak rate below 7 CFM is !
based upon the potential radiological consequence of the design basis accident (DBA) as follows:
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, If ths DBA were to occur a fraction of the fuel radioactivity inventory would be released into the containment air. There would be a brief positive containment pressure spike. Quickly, however, the 6000 CFM powerhouse fan would draw the containment back to a negative 1/2 inch pressure. The air would be drawn through I
the absolute and charcoal filtered 6" emergency exhaust duct.
Once the containment was drawn down to the negative 1/2 inch control pressure, small amounts of air would have to be drawn out over an extended period of time. The amount which would have to be removed would be the amount which would leak inward into the containment at the negative 1/2 inch pressure. Therefore, the 7 CFM Tech. Spec. limit would determine the rate of release of contaminated air as a consequence of the DBA.
The FSAR indicates that the dose to the public as a result of the DBA releases would be minute. In 1974 the Commission granted a modification to our Technical Specifications to allow us to operate the reactor fuel to higher cumulative burnups. As a result the inventory of nuclides not in equilibrium at the lower burnups will increase.
Calculations accepted by the Commission at that time demonstrated that even at the higher bygnups the doses to the public will be on the order of 10- that which is allowed. This issue was again analyzed upon renewal of License R-77.
Therefore, at the measured leak rate of 19 CFM the potential dosestothepublicintheeventoftheoccurrenceofb*f DBA would have been on the order or 19/7 or 2.7 x 10~o o <
the limit.
Leak Rate Surveillance - Observability of the Problem I Potential sources contributing to a high building leak rate
- include the following
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- 1. Leakags of outside air past the airlock or truck door gaskets.
- 2. Leakage past small containment penetrations such as electrical or plumbing.
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- 3. Leakage past the large diameter (core inlet and outlet) containment penetrations.
- 4. Leakage around the outside of ventilation ducts which penetrate the containment walls.
- 5. Failure of one or more of the hydraulic isolation dampers to seal under accident conditions.
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'. . 6. Increases in the porosity of the concrete / paint of the containment surfaces.
- 7. Cracking of the containment walls, roof, or floor.
- 8. Leakage from compressed air systems into containment.
(Including airlock and truck door gaskets, lab spigots, and ventilation and heating control air.)
Significant Leakage through the Truck Door or airlock gasket may be observed audibly,however small ones can not. The same holds true for small containment electrical and plumbing penetrations.
The isolation dampers are normally open, but close in response-to manual or automatic (high radiation) trip. It is not possible to visually observe the position of the butterfly . However BMRC has added a mark to the position of the rotating shaft for the 36" damper and is not likely due to the strong sturdy construction that-the butterfly would misalign with the shaft.
The ability to discern cracks in the containment wall would depend on the size and location of the crack. It is likely that crack large enough to cause significant leakage would arise only from major or catastrophic events.
There is no way to observe the porosity of the paint / concrete surfaces. However the contaminant has been painted infrequently and no obvious trends in the leak rate have been observed.
Leakage from compressed air systems into the containment could also be difficult to detect since many components are not readily assessable.
Ambient noise levels in the containment are high and would hamper detection of small leaks on a day to day basis.
The current Technical Specification limit of 7 CFM is very small and in light of the numerous potential sources of leakage, it is not readily observable. Many larger leaks such as gasket or compressed air leaks may be detectable by observation (listening). However, the only reliable way to determine conformance with the Technical Specification limit would be to perform the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leak test. It is not practicable to perform this test on a frequent basis.
In light of the abcVe, and supported by the small radiological consequences an increuse of the Technical Specification to 30 CFM was requested on December 20, 1986. The Commission requested additional information in support of the Technical Specification change on 4/14/89. BMRC is currently preparing the information requested.
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". Failure of the 36" Roof Damner On 4/24/89 a positive pressure was placed on the containment (reactor shut down) and the operations staff searched for signs of leakage in the various containment penetrations. Minor leaks were discovered in two of the airlock gaskets. These leaks have been patched. Replacement gaskets were ordered in March 1989 and delivery is expected around mid May.
The BMRC Operations Manager descended into the 36" exhaust vent to check the seal of the isolation damper. There was noticeable air flow around the damper. A gap of about 1/32 of an inch thickness was observed between the butterfly plate and the rubber seal, over an arc of about 1.5 feet. Please refer to figure 1 .
Initially it was believed that the poor seal was a result of wearing and/or aging of the thick rubber seal (original equipment) a new seal was ordered. Corrosion and wear of the butterfly plate appeared to be minimal.
On 4/25/89 an access port was cut into the duct wall to allow better observation and to facilitate replacement of the seal.
Once the duct was opened it was determined that the butterfly did not exhibit proper rotational alignment. Instead of rotating in a 90 degree arc from the vertical to a horizontal orientation, the butterfly was approximately 10 degrees out of time.
On 4/26/89 the alignment was adjusted to the correct orientation.
The damper seal was tested by the soap bubble method with the building at a positive pressure of approximately 12 inches.
Minor leakage near the shaft / seal penetration was observed.
Leakage at this point is probably minimal at the lower design pressure of -1/2 inch.
A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leak test was initiated at 4:15 P.M.. The leak rate determined by this test was 4.7 CFM which is below the Technical Specification Limit.
Safety Committee Review On 4/27/89 the Nuclear Safety Committee convened and reviewed a preliminary report of the " Reportable Occurrence". The committee reviewed the available information and corrective acticns taken to date. At the time the committee convened the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leak test was not yet complete, however extrapolating the leak containment leakage to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> indicated that the test would likely be positive.
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. Bag d on ths cbova the Committeo agr :d that contingsnt upon successful completion of the leak test the reactor would be in compliance with Technical Specifications and could be safely restarted. It was further agreed that the leak rate would be confirmed after installation of the new airlock gaskets. It was also agreed until further notice to verify the rotational position of the 36" butterfly by observation of the shaft / mark quarterly.
The committee further reviewed the issue of the monitor by-pass air flow as outlined below.
At the successful conclusion of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leak test the reactor was returned to normal service.
Design Error in the Buildina Air Radiation Monitor I
l A design error in the Building Air monitor system was identified in the course of investigating the high containment leak rate.
Air is drawn from the exhaust duct downstream of the filter and damper systems. It passes through a fixed filter particulate monitor and then through a gas monitor. The monitor stream then passes through the " Windjammer" pump and is returned to the duct slightly downstream of the sampling point. The sampling is not isokinetic but is conservatively performed at a velocity higher than duct average.
The Windjammer is a positive displacement pump. Flow, therefore, cannot be adjusted by throttling the loop. Instead a bypass stream is drawn into the system. (Please refer to Figure 2 )
This bypass air is drawn from the fan room. This is not acceptable for two reasons. First this flow of air constitutes a release of unfiltered air downstream of the sampling tube. More l importantly in the event of the DBA or other serious accident l generating airborne radioactivity in the containment, the monitor would continue to pump air out of the containment. This leak would be unfiltered and released from the containment roof.
Analysis of the DBA and the impact of the leak rate therein
{ assumes a filtered release through the 50 meter exhaust stack.
Within the context of the FSAR, therefore, the releases associated with the monitor bypass flow would be significant.
The solution to this problem is to draw the bypass flow from the exhaust duct. In this manner in the event of an accident the i monitor would simply circulate outside (effluent) air through the system and not containment air.
The Nuclear Safety Ccmmittee agreed with BNRC management that it would be appropriate to replumb the bypass to draw air from the duct. This modification was made before the reactor was restarted.
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... . Actions to Prevent Re-Occurrence As outlined above actions taken to prevent re-occurrence of the deviation from Technical Specifications are:
- 1. The leak rate measurement will be repeated after installation of replacement gaskets.
- 2. The orientation of the 36 inch damper shaft will be observed at least quarterly.
- 3. The monitor bypass line has been~repiped to draw exhaust air.
- 4. The design of the stack monitor has be evaluated and does not pose similar problems.
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