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{{#Wiki_filter: | {{#Wiki_filter:February 26, 1974 Re: Indian Point Unit No. 2 AEC Docket No. 50-247 Operating License DPR-26 A.O. 4-2-9 Mr. James P. O'Reilly, Director Regulatory Operations, Region I | ||
: u. S. Atomic Energy Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 | |||
' | |||
==Dear Mr. O'Reilly:== | ==Dear Mr. O'Reilly:== | ||
In accordance with the requirements of Section 6.12.2(a) of the Technical Specifications of Facility Operating License No. DPR-26, the following report is submitted! | In accordance with the requirements of Section 6.12.2(a) of the Technical Specifications of Facility Operating License No. | ||
In the course of perforrning | DPR-26, the following report is submitted! | ||
In the course of perforrning . periodic surveillance test PT-Mll, "Steam Line Pressure Analog .Channel Functional Test" on February 22, 1974, an inadvertent safety injection signal was generated which, by design, caused the accumulator discharge stop valves to open. At-the time of the occurrence, the reactor was in the cold shutdown condition with the Re~idual Heat Removal System in service and a reactor coolant pressure and temperature of 150 psig and ll50F respectively. Since the reactor coolant system was be-ing operated in the solid mode, opening of the accumulator valves pressurized the system to about 560 psig which was slightly in excess of the Technical Specification limit 'of 500 -psig for in-dicated temperatures at or below 2200F. The pressure was promptly reduced below the 500 psig limit by operator action. | |||
Since the reactor coolant system was ing operated in the solid mode, opening of the accumulator valves pressurized the system to about 560 psig which was slightly in excess of the Technical Specification limit 'of 500 -psig for dicated temperatures at or below 2200F. The pressure was promptly reduced below the 500 psig limit by operator action. There was no damage incurred to any system or component as a result of the pressure transient of this magnitude nor was there any reason to expect any.* . Mr. Anthony Fasano of your office was informed of this occurrence by Mr. John M. Makepeace on February 22, 1974. Very truly yours, . L.Uo . | There was no damage incurred to any system or component as a result of the pressure transient of this magnitude nor was there any reason to expect any.* | ||
12_ .. | . | ||
Mr. Anthony Fasano of your office was informed of this occurrence by Mr. John M. Makepeace on February 22, 1974. | |||
t , | |||
Very truly yours, . | |||
L.Uo. >J~,, 12_ 00~. . | |||
Warren R. Cobean, Jr * .Manager cc: John F. O'Leary Nuclear Power Generation Depart. | |||
l}} | |||
Revision as of 19:19, 29 October 2019
| ML17252A845 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/26/1974 |
| From: | Cobean W Consolidated Edison Co of New York |
| To: | O'Reilly J US Atomic Energy Commission (AEC) |
| References | |
| AO 4-2-9 | |
| Download: ML17252A845 (1) | |
Text
February 26, 1974 Re: Indian Point Unit No. 2 AEC Docket No. 50-247 Operating License DPR-26 A.O. 4-2-9 Mr. James P. O'Reilly, Director Regulatory Operations, Region I
- u. S. Atomic Energy Commission 631 Park Avenue King of Prussia, Pennsylvania 19406
'
Dear Mr. O'Reilly:
In accordance with the requirements of Section 6.12.2(a) of the Technical Specifications of Facility Operating License No.
DPR-26, the following report is submitted!
In the course of perforrning . periodic surveillance test PT-Mll, "Steam Line Pressure Analog .Channel Functional Test" on February 22, 1974, an inadvertent safety injection signal was generated which, by design, caused the accumulator discharge stop valves to open. At-the time of the occurrence, the reactor was in the cold shutdown condition with the Re~idual Heat Removal System in service and a reactor coolant pressure and temperature of 150 psig and ll50F respectively. Since the reactor coolant system was be-ing operated in the solid mode, opening of the accumulator valves pressurized the system to about 560 psig which was slightly in excess of the Technical Specification limit 'of 500 -psig for in-dicated temperatures at or below 2200F. The pressure was promptly reduced below the 500 psig limit by operator action.
There was no damage incurred to any system or component as a result of the pressure transient of this magnitude nor was there any reason to expect any.*
.
Mr. Anthony Fasano of your office was informed of this occurrence by Mr. John M. Makepeace on February 22, 1974.
t ,
Very truly yours, .
L.Uo. >J~,, 12_ 00~. .
Warren R. Cobean, Jr * .Manager cc: John F. O'Leary Nuclear Power Generation Depart.
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