ML20137H865
| ML20137H865 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 11/27/1985 |
| From: | Randazza J Maine Yankee |
| To: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| MN-85-201, NUDOCS 8512020493 | |
| Download: ML20137H865 (8) | |
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MAIRE HARHEE MGMl0 POWER 00MPARUe EDISON DRNE AUGUSTA, MAINE C4336 (207) 623-3521 lI November 27, 1985 MN-85-201 United States Nuclear Regulatory Commission j
Office of Inspection and Enforcement Washington, D. C.
20555 Attention:
Mr. James M. Taylor Director, Inspection and Enforcement
References:
(a) License No. DPR-36 (Docket No. 50-309)
(b) USNRC Letter to MYAPCo dated August 28, 1985 (c) USNRC Letter to MYAPCo dated September 4,1985 (d) MYAPCo Letter to USNRC dated September 13, 1985 (MN-85-164)
(e) MYAPCo Letter to USNRC dated October 15, 1985 (MN-85-177)
(f) USNRC Letter to MYAPCo dated October 29, 1985
Subject:
Response to Notice of Violation and Proposed Civil Penal'ty Gentlemen:
Tais letter forwards our response to Reference (f).
The events are described accurately and fairly in the Inspection reports, References (b) and (c).
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We consider the violations to be serious since they resulted in a possible degradation of a safety system and have implemented and completed a fourteen point corrective action program as described in References (d and e).
The corrective actions taken will reduce the probability of such errors in the future and are consistent with those recommended in Reference (f).
The comprehensive corrective action program included a review of all plant design changes implemented since 1973 and verification of all instrumentation root valve positions to assure that similar problems did not exist at the facility. No other errors were identified as a result of the design review and no other valves were found to be mispositioned indicating that these were isolated occurrences.
The instrumentation involved provides two safety functions. It is a sub-system of the reactor protective system serving to trip the reactor in the unlikely event of an instantaneous and catastropic rupture of a main steam line. It also serves to close off feedwater flow mitigating the reactor cooldown transient that would result from such an event 8512020493 851127 PDR ADOCK 05000309 0
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l United States Nuclear Regulatory Commission Page Two Attention:
Mr. James M. Taylor MN-85-201 We believe the categorization of the violation as Severity Level II may be inappropriate. It appears to be based in part, on the misimpressions that the '
instrumentation was completely inoperable, and that a complete loss of a primary trip function existed. For the reasons set forth below, we request that the categorization be reconsidered.
During our presentation at the Enforcement Conference on September 9,1985 and at the technical meeting on September 20, 1985, we stated that the instrumentation involved correctly indicated and responded to changes in the monitored plant parameters (steam generator pressures) throughout the fuel cycle. Therefore, we were unaware that any problem existed. We stated that Channel A was considered to be incapable of providing its reactor trip function because of the lead commoning design error but that its feed pump trip function was unaffected. We also stated that while the responsiveness of Channels B, C & D may have been affected by the valve mispositioning, the resultant delay (if any) would not have been significant and that there was a high probability that all of the expected automatic protective actions provided by the sub-system would have occurred in the event of a design basis accident.
We also presented the results of our evaluatilon of hypothetical case of a complete failure of the low steam generator pressure trip sub-system to function in the event of a design basis accident. In such a case other sub-systems of the reactor protective system would have tripped the reactor performing the necessary safety function. While this is acknowledged in the notice of violation, our presentation, heavily weighted to providing assurance that reactor trip protection existed, may have inadvertently, drawn some to the conclusion that the complete sub-system was inoperable.
The other protective feature provided by the instrumentation is a trip of the feedwater system to mitigate the severity of the inevitable cooldown transient consequences of a design basis steam line break.
This function was not part of the original plant design or license basis. It was installed in 1981 to provide additional assurance of feedwater flow reduction. The operability of the original cooldown mitigation equipment was not affected by the violation.
Even if the steam generator pressure channels were assumed not to function at all in response to a design basis event, the resultant cooldown transient on the reactor pressure vessel would not have been more severe than that calculated in the latest licensing case. The attached plot of cold leg temperature vs. time (Attachment A) illustrates this point. The licensing case and a best estimate case without the trip of the feedwater system, are not significantly different.
Attachment B contains our specific response to the two violations.
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United States Nuclear Regulatory Commission Page Three Attention:
Mr. James M. Taylor MN-85-201 j
Based on the foregoing we do not believe that the problems resulted in "the complete loss of a primary trip function" as expressed in Reference (f) nor did they result in "a system not being able to perform its intended safety function" as expressed in 10 CFR Part 2 Appendix C (1985). Therefore, we request that the category level of the violations be reconsidered.
We are requesting deferral of the civil penalty pending outcome of our request for reconsideration of the category level of the violations (Attachment C).
We are not specifically requesting a mitigation of the civil penalty, however you may deem it appropriate if the category level of the violation is modified.
Very truly yours, MA1NE YANKEE ATOMIC POWER COMPANY
@Q.D-Ju John B. Randazza Executive Vice President JBR/bjp
Enclosure:
cc: Dr. Thomas E. Murley Mr. Cornelius F. Holden Mr. Patrick Sears STATE OF MAINE Then personally appeared before me, John B. Randazza, who being duly sworn did state that he is Executive Vice President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and file the foregoing submittal in the name and on behalf of Maine Yankee Atomic Power Company, and that the statements therein are true to the best of his knowledge and belief.
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ATTACHMENT A
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Attachment B MAINE YANKEE RESPONSE TO NOTICE OF VIOLATION 9
Violation A Technical Specification Limiting Condition for Operation (LCO) 3.9, and Table 3.9.1, Instrument Operating Requirements for the Reactor Protective System, requires that whenever the reactor is in power operation.. a minimum of three of the Reactor Protective System channels must be operable for low Steam Generator Pressure.
Contrary to the above, from June 22, 1984 until August 7,1985, with the reactor in power operations, all four channels of the Reactor Protective System for each of the three steam generators were inoperab]u for low Steam Generator Pressure.
Violation B Technical Specification LCO 3.22, Feedwater Trip System, requires that whenever the reactor coolant boron concentration is less than that required for hot shutdown, the feedwater trip system shall be operable to assure automatic shutdown of all main feedwater pumps, automatic closure-of all main feedwater valves, and automatic closure of all auxiliary feedwater valves.
Contrary to the above, from June 20, 1984 until August 7,1985, with the reactor coolant boron concentration less than that required for hot shutdown, three of the four channels for the Feedwater Trip System for each of the three steam generators were inoperable for low Steam Generator Fressure.
Response
We admit the violations occurred and believe the violations to be serious since they could have resulted in a degradation of a safety protection system.
During the referenced period all four channels (twelve indicators) of the steam generator low pressure trip sub-system indicated correctly and responded appropriately. Therefore, it was not apparent that any problem existed.
The cause of the mispositioned instrument root valves was a lack of clarity in a test procedure which contributed to operator error. The cause of the instrumentation commoning design error was a failure of the engineer to thoroughly search out possible interactive circuits coupled with an inadequate design review.
When each of the two problems were detected, they were promptly and accurately reported to the NRC. Corrective action was taken promptly and comprehensive l
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. measures were implemented to ensure similar problems did not exist and to prevent recurrence. They are described in Reference (d) and (e).
Full compliance was achieved prior to the reactor startup on October 24, isd5. -
The extensive design review and valve position verifications performed as part of the corrective measures, uncovered no similar problems. Therefore, these were isolated occurrences, primarily the result of personnel errors. They were not the result of significant programmatic deficiencies or other broad underlying management control problems.
For the reasons set forth below, we believe the categorization of these violations as Severity Level II is inappropriate and request that it be reconsidered.
I The steam generator low pressure trip channels constitute a sub-system of the reactor protection system designed primarily to assure a reactor trip in the unlikely event of a catastrophic steam line break. The sub-system also functions to trip the feedwater system mitigating the resultant overcooling transient.
We concur that Channel A was inoperable to provide a reactor trip function.
Because of the design errot, Channel A would have indicated the average pressure of all three steam generators, therefore, would not have responded adequately to cause a reactor trip as a result 6f low pressure in one steam generator. However, since the design error did not affect the feedwater trip portions of the circuitry, Channel A was operable for that function.
Channels B, C, and D were responsive to small changes in steam generator pressure during operation. There is no reason to believe they would not have responded to the large change in steam generator pressure that would occur in the event of a steam line break.
Based on the foregoing, we do not believe that the problems resulted in "the complete loss of a primary trip function" as expressed in Reference (f).
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Additionally, Title 10 of the Code of Federal Regulations, in Appendix C to Part 2 provides guidance on the categorization of the severity of violations.
A Severity II violation is defined, in part, to exist in the event of "(a) system designed to prevent or mitigate serious safety events not being able to perform its intended safety function." The reactor protection system is the system designed to trip the reactor in the event of design basis events.
Even if it is assumed that the low steam generator pressure trip sub-system would not function, other sub-systems of the reactor protective systems would have tripped the reactor, performing its intended safety function, in the event of any design basis accident including a catastrophic steam line break.
Furthermore, the overcooling transient on the reactor vessel would not have been more severe than that calculated in the latest licensing basis case.
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. In Summary:
. We identified, promptly reported, and corrected the problems.
. Wc instituted and completed a comprehensive corrective action program to reduce the probablity of reoccurrence,
. Despite the violations, we believe the steam generator low pressure subsystem channels would have functioned to trip the reactor and mitigate the cooling of the reactor vessel.
. We believe the categorization of the violation as Severity Level II is
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inappropriate and request that it be reconsidered.
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ATTACHMENT C Response to the Proposed Civil Penalty Maine Yankee requests deferral of the civil penalty pending outcome of out request for reconsideration of the category level of the violations. We are not requesting mitigation of the civil penalty, however, mitigation of the civil penalty may be appropriate if the category level of the violation is modified in accordance with our request.
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