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{{#Wiki_filter:February 26, 1974 Re: | {{#Wiki_filter:Mr. James P. O'Reilly, Director Regulatory Operations, Region I | ||
: u. S. Atomic Energy Commission 631 Park Avenue February 26, 1974 Re: | |||
Indian Point Unit No. 2 AEC Docket No. 50-247 Operating License DPR-26 A.O. 4-2-9 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. | 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! | 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. | 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.* | 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. | 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.. >J~,, 12_ 00~ | |||
Very truly yours, | .... t, | ||
L.Uo. >J~,, 12_ 00~. . | Warren R. Cobean, Jr *.Manager Nuclear Power Generation Depart. | ||
Warren R. Cobean, Jr * .Manager cc: | cc: | ||
l}} | John F. O'Leary l}} | ||
Latest revision as of 11:27, 8 January 2025
| 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
Mr. James P. O'Reilly, Director Regulatory Operations, Region I
- u. S. Atomic Energy Commission 631 Park Avenue February 26, 1974 Re:
Indian Point Unit No. 2 AEC Docket No. 50-247 Operating License DPR-26 A.O. 4-2-9 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.
Very truly yours, L.Uo.. >J~,, 12_ 00~
.... t,
Warren R. Cobean, Jr *.Manager Nuclear Power Generation Depart.
cc:
John F. O'Leary l