ML17252A884
| ML17252A884 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 11/14/1973 |
| From: | William Cahill Consolidated Edison Co of New York |
| To: | O'Reilly J US Atomic Energy Commission (AEC) |
| References | |
| Download: ML17252A884 (3) | |
Text
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William J. Cahill. jr.
- Consolidated Edison Company of New York. Inc. S 4 Irving Place. New York. NY 10003 Telephone (212) 460-3819 Regulatory Docke_t File November 14, 1973 Re Mr. James P. O'Reilly, Director Directorate of Regulatory Operations U. s. Atomic Energy Commission Region 1 631 Park Avenue King of Prussia, Penn.
19406 Dear Mr. q'Reilly This letter is to inform you of an incident which Indian Point Unit No. 2 on November 13, 1973 between 7:40 a.m.
and 10:00 a.m., and the preliminary results of our investigation to date.
The plant was being prepared for resumption of operation following an outage for maintenance.
At approximately 7:40 a.m., the reactor was critical and at approximately 7% power, all preparations were being made for synchronization of the main generator to the system.
At this time a turbine trip occurred due to high feedwater in the Steam Generator No. 23 which was apparently unrelated to the subsequent incident.
This trip resulted in a shutdown of the operating main boiler feed pump which in turn caused the level in all four steam generators to decrease.
At 7:45 a.m., there was a reactor trip due to low level in No. 21 steam generator.
Shortly thereafter, the feedwater line to Stearrt*Generator No. 22 experienced shaking accompanied by a load noise.
Since the turbine had previously tripped*~ feedwater was being supplied to all four steam generators via the auxiliary feedwater system.
The feed-water lines to the other. steam generators (Nos. 21*, 23 and 24) appeared to be quiet and did not experience the movement similar to that observed iri the line to Steam Generator No. 22.
The operators noted that proper steam generator water level could not be maintained in Steam Generator No. 22, and that there was an apparent leak inside containment, as indicated by increasing water level.in the containment sump.
Radiation, temperature and pressure having been determined to be at acceptable levels, an inspection party entered containment.
Steam and water was observed coming from the area of the feedwater lines n*ear the containment penetration above Elevation 46' and* there 8322
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James P. O'Reilly Atomic Energy Commission 'i November 14, 1973 Re *Indian Point Unit No. 2 AEC Docket No. 50-247 License No. DPR-26 was water on the containment floor.
The plant operators immediately began cooldown of the primary system to the cold shutdown condition using auxiliary boiler feed flow to Steam Generators 21, 23 and
- 24.
Feedwater flow to N,o. 22 steam generator was secured.
At approximately 1 p.m., Steam Generator No. 22 steam pressure reached zero psig and access was then possible to the area of containment, where steam had previously been* observed.
A visual examination identified a crack in the No. 22 steam generator feedwater line located above Elevation 46' and adjacent to the containment penetration.. Further examination determined that
- the crack extended for approximately 180° around one side of the line at and adjacent to the weld attaching the pipe to the end plate in the containment penetration pipe sleeve.
The containment penetration pressurization system for this penetration rem.ained leak-tight as established by check of the monitoring system following the incident.
The plant has been brought to a cold shutdown condition and a thorough investigation is.underway to determine the cause of the event, the damage resulting, repair actions necessary and action required to prevent recurrence of the incident.
Mr. Oberg of your office was at the plant on November 13, 1973 and was informed by plant personnel of the incident and accompanied them on the inspection of the affected areas.
We wi'll provide you with supplemental information concerning this incident as the results of our investigation become available.
This incident did not involve any apparent increase in containment pressure, nor any anomalies in primary system temperature, pressure and reactivity, and no radiation was released either to containment or to the environment.
The reactor was in a shutdown condition at the time of the incident; and was brought to a cold shutdown condition in an orderly manner following the incident.
The RHR system and three steam generators are available for maintaining the reactor in a safe shutdown condition and for removal of decay heat from the core.
md Very truly yours
&~:?l,rlt:~
William J. Cahill, Jr.
Vice President
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Copy to John F. O'Leary, Director Directorate 0£ Licensing U. s. Atomic Energy Commission Washington, D. C.
20545
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