ML20086L732
| ML20086L732 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/15/1974 |
| From: | Cooney M PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20086L737 | List: |
| References | |
| AO-74-8, NUDOCS 8402080486 | |
| Download: ML20086L732 (3) | |
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March 15, 1974 Mr. A. Giambusso
' Deputy Director of Reactor Projects United States Atomic Energy Commission Directorate of Licensing 50-2,77 Washington, D C 20545
Dear Mr. Giambusso:
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Subject:
Abnormal Occurrence
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The fol, lowing occurrence was reported to Mr. R. A. Feil, A.E.C.
tj Region I Regulatory Operacions Of fice on March 8,1974. Written notifi-
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cation was made to tir. James P. O'Reilly, Region I Regulatory Operations Office on March 8, 1974.
In accordance with Section 6 7 2.A of the Technical Specifications, Appendix A of DPR-44 fer Unit 2 Peach Bottom Atomic Power Station, the following report is being' submitted to ' the Directorate of Licensing as an Abnormal Occurrence.'
heference:
License Number DPR-44, Amendment Number 1, Technical Specification Reference 3 6.A.2 Report No.:
50-277-74-8 Report Date:
3/8/74
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Occurrence Date:
3/6/74 Facility:
Peach. Bottom Atomic Power Station
' Unit 2
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R.D. 1, Delta, Pennsylvania ~
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Identification of Occu rence:
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. f Followinga[mainsteamlineisolationtest, portions of the reactor vessel shell temperatures decayed to less than 185 F while reactor pressure remained above 250 psig. This exceeds a limiting con-dition for operation as stated in Technical Specification paragraph, fj,
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Conditions Prior to Occurrence:
Reactor shutdown following a main steam isolati.on test 4hich Jresulted in tripping of both re. circulation pumps. React r water / clean-f>' system and reactor core isolation system in service.
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v Mr. A. Giambusso March 15, 1974 Page 2 Description of Occurrence:
Following a start-up test which isolated the main steam lines with the plant at a nominal 25% power, the reactor was shutdown and primary system conditions were stabilized with both recirculation pumps out of service and the control rod drive pumps out of service. The reactor water clean-up system was in service on full recirculation.
-Reactor vessel icvel was maintained by the reactor core isolation cool-ing system. This system also maintained reactor vessel pressure essen-tially constant at approximately 400 psig by taking a small flow from the upper part of the vessel. These conditions, with no forced circu-lation through the bottom vessel head, the bottom vessel shell cooled
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to approximately 125 F.
This condition exceeds the limi ting condition for operation which states: "The reactor vessel shall not be pressur-ized above 250 psig unless the reactor vessel temperatures are equal tc or greater than 185 F if fuel is in the reactor vessel."
Designation of Apparent Cause of Occurrence:
The main steam isolation start-up test results in innediate closure of all main steam isolation valves which in turn cause a reactor scram. With the loss of steam of the Turbine / Generator, the Turbine / Generator must be tripped which in turn causes fast transfer of the auxiliary busses and tripping of both recirculation pumps. All plant systems operated in accordance with design during this start-up test. The main steam isolation test was followed innediately by an-other start-up test which demonstrates the ability to control reactor vessel level fran outside the Control Room. During this demonstration, operations which must be performed from the Control Room associated with the reactor systems were delayed until the Shutdown From Outside The Control Room Start-Up Test was completed. The delay in restarting the recirculation pumps following the isolation permitted the vessel temperatures to diverge to the point where the recirculation pumps could not bc restarted without violating a Technical Specification.
Additionally, the reactor vessel pressure was not decreased below 250 psig when the vessel temperature decayed to 185 F.
Analysis of Occurrence:
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The nucicar steam supply system vendor has verified that the vessel was not in any danger of crack initiation or other damage due to high stress. This assurance is based upon preparation of pressurization temperature curves in accordance with current ASME Boiler and Pressure Vessel Code Section III Appendix G procedures. These pressurization temperature curves represent limits which are more consistent with re-cent investigation in the area of control of brittle fracture.
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the transient, vessel wall temperatures did not drop below the pressuri-zation temperature curves. Also, because of the very slow nature of the cooldown and relatively low pressures (less severe than postulated events included in reactor pressure vessel stress and fatigue analysis), the vessel was in no danger of crack initiation or other damage due to high stress.
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O Mr. A. Giambusso March 15,1974 Page 3 Corrective Action:
Instructions have been issued to return the recirculation pumps to service as soon as possible following any plant transients.
If this cannot be accomplished, the reactor relief valves 'shall be utilized to depressurize the reactor to less than 250 psig before vessel shell temperatures can decay to less than 185cp, Very truly yours, I.J C o' M
Asst / Gent. Supt.
Generation Division cc:
Mr. J. P. O'Reilly Director, Region i United States Atomic Energy Commission 631 Park Avenue King of Prussia, PA 19406 s
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