ML19323E997
| ML19323E997 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/08/1980 |
| From: | Richard Bright FLORIDA POWER CORP. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8005280345 | |
| Download: ML19323E997 (3) | |
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Florida Power CO A PO R A Y DO N May 8, 1980 File:
3-0-3-a-4 Mr. J. P. O'Reilly Di rector U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Suite 3100 101 Marietta Street Atlanta, GA 30303
Subject:
Crystal River Unit 3 Docket No. 50-302 5
Operating License No. DPR-72 I.E.Bulletin 80-04 Analysis of a Power Main Steam Line Break
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with Continued Feedwater Addition j
Dear Mr. O'Reilly:
l Enclosed is our response to I.E.Bulletin 80-04.
Please contact this office if you require any additional discussion concerning our response.
Very truly yours, FLORIDA POWER CORPORATION R. M. Bright Acting Manager Niiclear Support Services RMBemhR02D3 cc:
NRC Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D.C. 20555 UFFICIAL COPY, y g g engrgI ffice 3201 Thirty fourth Street South e P.O Box 14o42, St Petersburg. Fionda 33733 e 813-866-5151 G
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i FPC RESPONSE TO I.E. BULLETIN 80-04 Response to Question 1 The containaent pressure response analysis relative to a main ; steam line break inside containment did not include the impact of runout from the Auxiliary Feedwater System nor the impact of other energy sources.
The runout flow of Auxiliary Feedwater and Main Feedwater were not considered, as CR-3 is provided with a steam line rupture matrix with the ability to detect and isolate the damage steam generator, following a main steam line rupture.
The main steam line rupture matrix detects a main steam line failure via main steam line header pressure.
Subsequent to a steam line rupture, both steam generators begin to blow down at the same rate.
The steam line rupture causes an increase in the heat transfer from the reactor coolant to the feedwater.
This initiates a cooldown of the reactor coolant system, which increases the reactor power, due to the large negative moderator coefficient, such that the reactor trips on high flux in approximately 6.5 seconds.
Reactor trip causes the turbine stop valves to close, isolating the unaffected steam generator on the steam side.
Loss of main steam line pres-sure will actuate the matrix thus initiating the closure of the main feedwater block valve, the feedwater low flow block, the feedwater pump suction valves and the auxiliary feedwater isola-tion valves.
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Following the isolation of the main steam lines the unaffected steam generator pressure recovers.
Upon pressure recovery of the unaffected steam generator, the prcssure switches activate the ICS to open the feedwater valve to permit feedwater flow to maintain the two-foot minimum level in the unaffected steam generator.
On the affected steam generator the feedwater isolation valves have been closed on low steam line pressure (no recovery), neither primary system pressure recovery nor a return of the reactor to power will reopen these valves. Hence, continued feedwater through the affected steam is precluded.
For detailed description of operation and considerations, refer to FSAR section 14, item 14.2.2.1 Steam Line Failure.
Since our February 26, 1980 transient at CR-3, Florida Power Corpora-tion has been evaluating the steam line rupture natrix system at CR-3.
As stated in our May 2,1980 submittal to the NRC, Florida Power has requested GAI and B&W to evaluate removing the isolation of the emer-gency feedwater valves from the rupture matrix.
Our concern is that this isolation of emergency feedwater may not be in the best interest of overall plant safety. This concern is based upon the higher 4*ea*$
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industry operating experience of actuation due to generator dry-out or RC system cooldown rather than a steam line rupture.
This evaluation of the containment pressure response and steam line rupture matrix will be submitted to the NRC for review upon completion.
If the evaluation supports not isolating EFW via the rupture matrix, a minor control change will remove values FWV-161 and FWV-162 from the rupture matrix actuation logic.
Response to Question 2 The steam line break analysis in the FSAR has been re. viewed and it has been determined that all potential water sources have been con-sidered as required by the licensing basis assumptions. Therefore, no corrective action has been identified.
Response to Question 3 Per our response to Questions 1 and 2 above, Question 3 does not apply to Crystal River Unit 3.
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