ML20056G840
| ML20056G840 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/30/1993 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD931033, NUDOCS 9309070225 | |
| Download: ML20056G840 (4) | |
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NSD931033 August 30, 1993 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, DC 20555
Subject:
Information Requested During August 13,
- 1993, Inspection Report 93-17, Enforcement Conference Cooper Nuclear Station NRC Docket No. 50-298, License No. DPR-46 Gentlemen:
During the August 13, 1993, Enforcement Conference to discuss Inspection Report 93-17, held at the Region IV Offices in Arlington, Texas, Nebraska Public Power District (District) committed to provide additional information on several topics.
The following information is provided in accordance with 10CFR50.4(b) in response to this commitment.
1.
Did the Corrective Action Program Overview Group (CAPOG) review various operating experience type of documents such as General Electric Service Information Letters, NRC Information Notices,
etc.,
as a part of its review of past documentation?
The CAPOG review of past documentation did not include a systematic review of operating experience documents (e.g., NRC Information Notices).
The review was directed towards the traditional CNS currective action program documents such as Nonconformance Reports, Deficiency Reports, Maintenance Work Requests, Operability Determinations, Operability Evaluations, and Radiological Safety Incident Reports, as well as certain NRC-related correspondence, including Bulletins and Generic Letters. It should be noted that during the review of specific subjects by the CAPOG, some past operating experience document responses were examined; however, as previously stated, a systematic review was not performed.
In consideration of the subjects discussed at the Enforcement Conference, a review of selected documents received over the last two years will be performed.
2.
Regarding secondary containment testing, the District stated that the April 7, 1993, test verified that the March 11, 1993, test was acceptable; yet, CNS personnel increased the negative differential pressure of the Radwaste Building from 0.1 inches to 0.2 inches wg to pass the March test. Was there any other work done on buildings or their penetrations after March 11 and before the April 7,1993, test that nould affect the results? Why would CNS have been able to pass the test on March 11 without a negative 0.2 inches wg Radwaste building pressure?
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Document Control Desk August, 30, 1993 Page 2 of 4 i
On the morning of March 11, with standby gas treatment running, an effort i
as during the was made to establish the test " configuration" the same 1991 Outage. As an element of this troubleshooting effort, the system l
engineer (a member of the troubleshooting team) for the Heating, j
Ventilation and Air Conditioning (HVAC) system was dispatched to ensure j
that plant building differential pressures were in their normal bands.
The Turbine Building differential pressure was verified to be in the normal band.
However, the system engineer found the HVAC differential pressure for the Radwaste Building to be outside its normal band and, at about 0930 and with assistance from Operations personnel, adjusted it from -0.1 inches wg to approximately -0.20 inches wg.
The effect of the change in Radwaste Building differential pressure on Secondary containment differential pressure was immediately noticed by Operations personnel in the Control Room and communicated to the troubleshooting team leader (the Plant Engineering Supervisor.)
l I
While waiting for Technical Specifications required wind conditions (vind j
speed of 2 to 5 mph), the Plant Engineering Supervisor reviewed the implications of the observed change in Reactor Building differential pressure when the Radwaste Building differential pressure was altered.
A review of the USAR, Chapter XIV, was conducted to determine the design basis requirements for the ventilation systems under accident conditions.
It was concluded that no concerns existed with regard to the refueling e
accident, but that prior to startup the anomaly would have to be investigated further due to the potential that the Radwaste Building i
ventilation system may not be available for the Design Basis LOCA i
scenario. Acceptable meteorological conditions existed that evening and l
the test was satisfactorily completed at approximately 2024 hours0.0234 days <br />0.562 hours <br />0.00335 weeks <br />7.70132e-4 months <br /> on i
March 11, 1993.
l On April 7,
1993, subsequent to modifying (SORC MVR 93-1204 for EVR 93-021) the fuel pool cooling rupture seal drain line to install a l
loop seal, a test of Secondary Containment was performed. This test was l
to serve as an acceptance test for the loop seal installation. The flow l
through the empty loop seal was determined to be less than 150 cfm. With this leakage factored in, the Acceptance Test Summary and Conclusions determined that the test on March 11 did, in fact, verify operability of
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i Subsequently, investigation by the Enforcement Issues Investigation Team j
of maintenance performed on Secondary Containment between March 11 and j
April 7, revealed minor maintenance activities which should not have significantly contributed to differences in the pressure retaining boundary of the secondary containment. This investigation was conducted
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by review of MWRs, and review and discussion of all Deviation from Outage Guidelines (DOGS) which were generated for all maintenance activities affecting containment. It can only be concluded from the available data j
that Secondary Containment repair activities prior to the March 11, 1993, test sufficiently reduced in-leakage such that secondary containment limits were met without the loop seal Installed on March 11, 190~4 l
b Document' Control Desk August; 30, 1993 Page 3 of 4 3.
Regarding RHR-MOV-M017 and M018, has NPPD looked at temperature concerns as well as elevated pressure on downstream piping as a result of the leakage? What was the setpoint of the relief valves that would mitigate the pressure increase, how often are they checked, and has preventive maintenance been performed on them?
The design temperature of the RHR shutdown cooling suction piping is 350 degrees Fahrenheit.
The maximum measured temperature on the discharge side of the RHR-MOV-M017 valve body while the leak was present was approximately 150 degrees Fahrenheit. The normal temperature in the piping during initial shutdown cooling, operations is 200-250 degrees Fahrenheit. Therefore, no design parameters were exceeded and the piping integrity was not challenged due to the observed leak.
1 The relief valve downstream of the RHR-MOV-M017 and M018 valves is a Dresser Model 1970C with a relief setpoint of 150 psig and a minimum flow capacity of 35 gpm (as compared to the measured leak rate past the valves of 0.35 gpm). The relief valve was tested per PM 03693 until 1991, and it was then included in the IST Program, which assures testing at least once per five year cycle. It was last tested satisfactorily on March 29, f
1993.
4.
Concerning Secondary Containment testing data, resolve any apparent t
difference between the LER submitted for the leak rate test failure and the information presented by the Enforcement Issues Investigating Team regarding the conduct of leak rate testing on March 10, 1993.
l
-i NRC IR 93-17 (at page 5) indicates that an undocumented Secondary i
Containment leak rate test was conducted on March 10.
LER 93-011 discusses testing on March 10, but does not discuss documentation of the j
testing. The Enforcement Issues Investigation Team reviewed the Control 1
Room and Shift Supervisor Logs for that date and found no record of an
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official test.
Initial interviews of personnel could not definitively l
establish whether a test was conducted on March 10, 1993.
Because an entry into SP 6.3.10.8 would have been logged tor an official test, some i
personnel indicated that perhaps a lineup for troubleshooting purposes
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(which would not be logged) may have occurred.
As a result, the Team
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initially concluded that no " testing" was conducted on March 10, 1993.
i llowever, based upon subsequent interviews and a review of the Control Room strip chart for Reactor Building differential pressure, the Enforcement Issues Investigation Team now concludes that indeed a test attempted on the evening of March 10, but the test was apparently was performed without initiating SP 6.3.10.8.
Late in the evening on March 10 Standby Cas Treatment was started and the Control Room strip chart for Reactor Building differential pressure was annotated
" Containment leak test."
When test parameters revealed that the acceptance criteria could not be achieved, the test was apparently terminated.
In summary, LER 93-011, submitted on May 12, 1993, is accurate in its description of the testing performed on March 10, as clarified herein.
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i Document Control Desk August, 30, 1993 Page 4 of 4 S.
Regarding the valve lineups to support integrated leak rate rests (ILRTs),
a)
When was the last system valve lineup performed for the 11 / 02 2
analyzer?
The last valve lineup was performed on June 28, 1993, per CNS Procedure 2.2.60.A.
b) llave there been other previous instances noted where a valve was found to be mispositioned during the conduct of an integrated leak rate test?
A review of ILRTs conducted since 1976 revealed only two other instances of mispositioned valves and one case where a relief valve had been temporarily removed and a blank flange installed.
The mispositioned valves were found and correctly positioned either during pretest lineups, or during the performance of SP 6. 3.1. 3, ILRT, and did not affect the test results.
It should be noted that SP 6.3.1.3 contains a comprehensive listing of primary containment interfacing valves that are required to be aligned in preparation for the ILRT.
Should you have any further questions concerning this submittal, please contact ne.
Sin rely.
- k a-G\\R llorn Vice President - Nuclear GRil/ dis /ya cc:
NRC Regional Office Region IV Arlington, Texas NRC Resident Inspector Cooper Nuclear Station