ML20236Y214
| ML20236Y214 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/04/1998 |
| From: | Miller H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Tom Gurdziel AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 9808120001 | |
| Download: ML20236Y214 (5) | |
Text
UNITED STATES
.. ; 8 NUCLEAR REGULATORY COMMISSION o
j REGloN I 475 ALLENDALE HGAD
[g KING oF PRUSSIA, PENNSYLVANIA 19406-1415 August 4, 1998 Mr. Thomas Gurdziel 9 Twin Orchard Drive
. Oswego, NY 13126
SUBJECT:
YOUR LETTER TO CHAIRMAN JACKSON DATED JUNE 19,1998 1
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Dear Mr. Gurdziel:
I am responding to your letter to Chairman Jackson, dated June 19,1998, in which you expressed concern over the restart of the Vermont Yankee Nuclear Power Station after the June 9,1998, reactor vessel high water level turbine trip and reactor scram. I am also
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aware of your letter to Mr. Cowgill of my staff dated June 17,1998, regarding the same event. Mr. Cowgill will respond to that letter in separate correspondence from this office.
Your June 19,1998, letter expressed the belief that the Vermont Yankee Nuclear Power Station should not have been allowed to restart after the June 9 event. While the Nuclear Regulatory Commission (NRC) shared many of your concerns regarding the complexity of a
this event, following a thorough review of the event, we determined that there was no basis to prevent restart of Vermont Yarikee. The NRC was concerned by the circumstances surrounding this event as initially reported to the NRC. In response, we immediately dispatched the Senior Resident inspector to the site and assembled a Special Inspection Team to review the event. The Specialinspection Team arrived onsite later that same day and spent four days on,ite independently evaluating the equipment performance and operator response to the event, assessing the licensee's event investigation, as well as assessing the safety significance of the event, inspection Report 50-271/98-09 documents the findings of the N9C special team inspection. Mr. Cowgill will include a i
copy of this report with his letter.
When deciding whether to shutdown a facility or, ti !n this case, whether the NRC should take action to prevent the plant from restarting, the NRC looks at many different aspects of the facility's performance. There are specific conditions in the facility's operating license
. which, if appropriate, would require a shutdown and NRC approval prior to restart. An example would involve exceeding a safety limit. The NRC found that no safety limits were exceeded during this event. Further., while some plant equipment did not perform properly, the special inspection team concluded that all safety systems functioned properly in j.
response to this event. Beyond the criteria specified in the license, the NRC also reviews personnel performance, procedures, material condition,' and other contributors to overall
' plant performance. If wa anted, when significantly poor performance is detected, the NRC would act to ensure the facility would not operate until required improvements in performance were completed. In this case, our special inspection team did identify some performance issues involving procedure adequacy, control of emergency diesel generator settings and inadequate correctivo actions for deficiencies identified during protective relrey calibration..These concerns resulted in two violations of NRC requirements that the
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9808120001 9 gogo 4
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PDR ADOCK 05000271 l
Mr. Thomas Gurdziel 2
licensee corrected prior to plant restart. The NRC concluded overall that these concerns were not of sufficient significance to prevent plant restart.
In your June 19 letter, you raised concerns about several specific aspects of the June 9 event, in the attachment to this letter, I have addressed each concern and provided our assessment. Many of these issues are addressed in the referenced inspection report, providing additional detailed discussion of the equipment and operator performance.
In conclusion, we found that the plant transient resulting from this event was within the bounds of the plant design and that all safety equipment operated as expected. Licensee personnel performance in response to the event will be addressed in the referenced inspection report. Additionally, the NRC was in frequent contact with the Vermont Yankee Nuclear Power Corporation to ensure that all concerns identified by either the licensee's event review team or the NRC specialinspection team regarding procedure adequacy, equipment performance and operator performance were appropriately addressed and corrected prior to the restart of the facility.
I trust that I have been responsive to your concerns. If you have any other concerns about this matter, please contact Mr. Curtis Cowgill of my staff at (610) 337-5233.
Sincerely,
/ 4fsY H ert Miller Regional Administrator
Enclosure:
As stated I
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f L __-_. _
l Mr. Thomas Gurdziel 3
Distribution w/o encl:
J. Callan, EDO W. Travers, DEDE H. Thompson, DEDR P. Norry, DEDM J. Blaha, OEDO S. Burns, OGC
' K. Cyr, OGC S. Collins, NRR B. Boger, NRR B. McCabe, OEDO M. Callahan, OCA R. Bores, RI.
C. Cowgill, RI R. Summers, RI M. Oprendek, RI D. Screnci, R1 M. Fudge, RI EDO Reading File
~ PUBLIC ' e DNs. 50-271 CRC No. 98-0593 EDO GT98-405 This correspondence addresses policy issues previously resolved by the Commission, transmits factual information, or restates Commission policy.
DOCUMENT NAME: G:\\ BRANCH 5\\1-VY\\GURDGT.WPD Ts receive a copy of this document, Indicate in the box:
"C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N"=
No copy i
OFFICE Rl/DRP l
NRR Rl/DRS
%,WP Rl/RA
' 6 riehl "HMiller NAME
- CCowgill
- RCroteau "RConte DATE 07/07/98 07/08/98 07/10/98 107/13/98 07/13/98 OFFICE OEDO OCM NAME
- *JCallan
- SJackson '
j DATE-07/16/98 07/27/98 l
OFFICIAL RECORD COPY
- concurred on previous page
- concurred via teleconference I
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9 ENCLOSURE 1: ASSESSMENT OF CONCERNS "The feedwater flow control valve failed high." NRC assessment: The licensee investigated the 'A' Feedwater Regulating Valve's (FRV) failure to close and found a cap screw lodged in the valve internals preventing its closure. No similar foreign material was found in the 'B' FRV. The licensee was not able to ider.tify the source of this material. The special inspection team preliminarily concluded that this foreign material concern contributed to the complexity of the event, but that the plant operators responded acceptably to this problem. While we were concerned that the source of this foreign material was not found, we concluded that the licensee's investigation was thorough and that the licensee completed sufficient action to ensure that no other similar foreign materials were present in this section of the feedwater system.
"No feedwater isolation valves were provided, or even just one of two did not close."
NRC assessment: By design, the facility does have remotely-operated, feedwater header blocking valves. However, these valves are not designed to automatically close for any plant condition and would not normally be used by plant operators in response to a feedwater transient because of the designed slow response time for the valves to stroke. The operators properly responded to the transient, as required by their procedure, but since the 'A' FRV was stuck open as a result of the foreign material, they were unsuccessful in preventing the reactor vessel water level from reaching the setpoint for a turbine trip.
"With operator recovery action allowed, none of three feedwater trains could be started at reactor pressure after bein'g tripped on reactor high water. Attempting to start the feedwater pumps disables the offsite power to both the emergency gmerator supplied powerboards." NRC assessment: The issues concerning the inability to start the reactor feecl water pumps (RFP) and the resultant loss of normal power to the electrical buses are related. Subsequent to the reactor scram and as an expected response to the plant conditions, the 'C' RFP automatically started. However, the pump tripped off within seconds due to a mechanical failure in an associated minimum flow valve. Both the 'A' and 'B' RFPs automatically started simultaneously in response to the 'C' RFP trip. The NRC verified that this response was also an expected condition given the design of the associated controls for this system. However, this resulted in an overcurrent condition on certain on-site electrical buses that led to a loss of off-site power to the non-vital 4kV bus No.1 and the vital 4kV bus No. 3. The special
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inspection team preliminarily concluded that the licensee's procedures were deficient j
regarding operation of this system. The licensee corrected this procedure concern prio.
I to restart.
"With a valid start signal for both, only one emergency generator starts and loads."
NRC Assessment: Only the 'B' EDC received a start signal because its vital 4kV bus No. 3 deenergizei The 'B' EDG subsequently operated as designed. The 4kV vital bus associated with the 'A' EDG was never deenergized which meant the 'A' EDG was not required to start.
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2 "The AC power supplied by the emergency generator is inadequate to power one RPS uninterruptible motor generator set, and it is disconnected automatically." Ml3_Q assessment: In this case, the reactor protection system (RPS) motor generator (MG) set tripped on under frequency when a residual heat removal (RHR) pump was started on the bus powered by 'B' EDG. The design of the RPS MG set does not include an "uninterruptible" feature; however, it is designed to complete its safety function even with a loss of power. The NRC found that this equipment performed as it would be expected given its design. The fact that its power supply tripped during the June 9 event when operatcrs started the RHR pump had no adverse affect. Further, the RPS had completed its safety function when it initiated the automatic scram during the event. While the team found no concern with the design or performance of the RPS -
MG set, review of this issue led to the discovery of a related concern involving the adequacy of the licensee's test procedures for the on-site emergency power supply i
system. This matter is described in the referenced inspection report.
1 "The DC motor on the 'A' reactor protection system (RPS) motor generator set does not take over and power the set as it should." NRC assessment: The RPS MG sets are AC to AC devices which take 480 voit AC and reduce it to 118 volt AC. They do not i
have a DC power supply or motor that would power the MG set on loss of the feeder AC power. The inspection team found the MG set power supply functioned as designed.
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