ML17252A845: Difference between revisions
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{{#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 | {{#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 | : u. S. Atomic Energy Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 | ||
==Dear Mr. O'Reilly:== | ==Dear Mr. O'Reilly:== | ||
| Line 26: | Line 25: | ||
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. | ||
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Revision as of 11:04, 4 February 2020
| 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.
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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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