ENS 43369: Difference between revisions

v t e ENS Event
16:30 May 17, 2007
Title	Degraded Condition - Degradation of Principal Safety Barrier
Event Description	Under 10CFR50.72 (b)(3)(ii)(A), the following event has been determined to be reportable. This section concerns the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. This criteria was determined to apply to material problems that cause abnormal degradation of or stress upon a principal safety barrier (in this case, Reactor Coolant System). On 5/17/07, a Nine Mile Point Unit 1 shutdown was in progress for forced outage inspection and repairs of the Turbine Lube Oil system. Prior to the shutdown, a discrepancy was being experienced between Reactor level readings from channel 11 and 12 Reactor Level columns. A step change increase of 2 inches of Reactor level had occurred on May 6th, 2007 on Reactor Level column 12. This change was coincident with an increase of 18 [degrees] F of the reference leg and 2 [degrees] F decrease of the variable leg. The level difference was then 5 inches with a Technical Specification limit of 6 inches based on current Drywell temperature. The condition was documented in the Corrective Action Program, compensatory actions were taken, and an Integrated Response Team was formed to investigate and determine the cause. Calculations by engineering determined a leak of 4 drops per minute could explain the observed indications. During the initial drywell entry a visual examination was performed on the reference leg instrumentation line associated with Reactor Level Column 12. The piping was visually examined for any evidence of leakage. The exam was performed from the bottom of the condensing pot piping to the drywell shell penetration. The examination revealed a wet area at the bottom of a 1.0 inch 90 degree pipe elbow at an associated socket weld. The weld had one drop approximately 1.0 inch from the toe of the welds on the bottom of the pipe. The drop was wiped away with a leather gloved hand, which did show evidence of leakage. After waiting approx. 15 minutes, no additional leakage was observed. Reactor pressure had been lowered to approximately 400 psig at the time of the inspection due to the Reactor shutdown in progress. The piping, welds and Reactor level column above were closely examined and no other wetness or evidence of leakage was found. Also an exam was performed below pipe weld 012 (surrounding areas) and again no evidence of leakage could be found. No staining or other discoloration could be seen. The observations were documented in the Corrective Action Program. The licensee also provided the following information: Location of the Leak: 12 Yarway Reactor Level Reference Leg Leak Rate: 1 drop per minute T.S. Limits: 5 gallons per minute Sudden or Long-term Development: Sudden Leak Start Date: 5-06-2007 Time: 08:10 a.m. EDT Coolant Activity: 2.02 E-3 microcuries/ml Secondary: N/A List of Safety Related Equipment Not Operational: None The licensee has notified the NRC Resident Inspector. * * UPDATE AT 1705 EDT ON 06/26/07 FROM BRIAN FINCH TO J. ROTTON * * * The licensee is retracting this report based on the following: Following plant shutdown the piping segment from the 12 Yarway Reactor Water Level Reference Leg initially believed to be leaking was removed from the station for examination and testing. The piping segment was subjected to a 1600 psig nitrogen pressure test for approximately one half hour with no indications of leakage. Additional nondestructive examinations were performed which also did not show evidence of through-wall indications. Based on the exploratory nondestructive examination and leak testing completed in the field and on the removed piping segment, it was determined that there was no actual leakage in the drywell associated with the 12 Yarway Reactor Water Level Reference Leg piping. Further leak testing identified a 5 to 6 drop-per-minute leak from the drain valves associated with the 12 Yarway Reactor Water Level Reference Leg. Based on engineering analysis, this amount of leakage through the drain valve assemblies accounted for the level instrumentation differences noted between channel 11 and 12 Reactor Water Level columns. Following maintenance and repair activities, in-situ leak checks were performed on the reference leg piping in the drywell and the associated instrument rooms outside the drywell during plant startup on 5/22/2007 at approximately 900 psi with no leakage identified. Due to the leakage being identified in the valve assemblies for the level instrumentation and not in the piping inside the drywell, this issue does not represent a reportable condition under 10CFR50.72 and therefore is being retracted. The drain valve has been repaired. The licensee informed the NRC Resident Inspector. Notified R1DO(Powell).

ENS 43369 +/-
Where
Nine Mile Point New York (NRC Region 1)
Reporting

10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
Time - Person (Reporting Time:+3.88 h0.162 days <br />0.0231 weeks <br />0.00531 months <br />)
Opened:	Matthew Minnick 20:23 May 17, 2007
NRC Officer:	Gerry Waig
Last Updated:	Jun 26, 2007
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v • d • e

v t e Emergency Notification System
Nine Mile Point with 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
WEEKMONTHYEAR2023-03-18T18:10:00 [Table view]
[ENS list for Nine Mile Point]

@@ Line 16: / Line 16: @@
 | event date = 05/17/2007 12:30 EDT
 | last update date = 06/26/2007
-| title = Degraded Condition - Degradation Of Principal Safety Barrier
+| title = Degraded Condition - Degradation of Principal Safety Barrier
 | event text = Under 10CFR50.72 (b)(3)(ii)(A), the following event has been determined to be reportable. This section concerns the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. This criteria was determined to apply to material problems that cause abnormal degradation of or stress upon a principal safety barrier (in this case, Reactor Coolant System).
 On 5/17/07, a Nine Mile Point Unit 1 shutdown was in progress for forced outage inspection and repairs of the Turbine Lube Oil system. Prior to the shutdown, a discrepancy was being experienced between Reactor level readings from channel 11 and 12 Reactor Level columns. A step change increase of 2 inches of Reactor level had occurred on May 6th, 2007 on Reactor Level column 12. This change was coincident with an increase of 18 [degrees] F of the reference leg and 2 [degrees] F decrease of the variable leg. The level difference was then 5 inches with a Technical Specification limit of 6 inches based on current Drywell temperature. The condition was documented in the Corrective Action Program, compensatory actions were taken, and an Integrated Response Team was formed to investigate and determine the cause. Calculations by engineering determined a leak of 4 drops per minute could explain the observed indications.

v t e Unit 1
GE-2
Reactor critical	Not Critical
Scram	No
Before	Cold Shutdown (0 %)
After	Cold Shutdown (0 %)

ENS 43369: Difference between revisions

Latest revision as of 22:16, 1 March 2018

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ENS 43369: Difference between revisions

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