ML19323F250
| ML19323F250 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/22/1980 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0667, RTR-NUREG-667 3-03-A-3, 3-3-A-3, NUDOCS 8005280697 | |
| Download: ML19323F250 (5) | |
Text
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&'fff[hk Mp Florida Power CO R PO R A T 40 M May 22, 1980 File: 3-0-3-a-3 I
Mr. Robert W. Reid
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Chief Operating Reactors Branch #4 Division of Operating Reactors U.S. Nuclear Regulatory Comission Washington, DC 20555 Subj ect: Crystal River Unit No. 3 Docket No. 50-302 Operating License No. DPR-72 NRC Letter Dated May 2,1980 Requesting Additional Information Restart After 2/26/80 Incident
Dear Mr. Reid:
Part II of the subject letter, requesting additional information to pre-pare Safety Evaluation to enable restart of Crystal River Unit 3, iden-tified areas not yet addressed by Florida Power Corporation with submit-tal of March 12, 1980, concerning the February 26, 1980, event at CR-3.
This letter responds to Part II questions of the subject letter.
Question II.2:
"...... Identi fy and describe the plant modifications to be made to reduce the number of main alarms to a manageable size for the reactor operators to handle.
Provide your schedule and major milestones for the completion of these modifications."
Response
i We have discussed this question (and Questions II.3, II.4, 11.5, 11.8, II.9, II.10, and II.11) and our current manpower limitations in. preparing for startup and responding to imedi-ate concerns with Messrs. Fairtile and Erickson on May 8, 1980.
We will respond to these question; when we provide re-sponses to NUREG-0667.
Both Mr. Fairtile and Mr. Erickson have concurred with this schedule in the telephone conversa-tion of May 8, 1980.
U n 05280 g General Office 32o1 inirty-fourtn street soutn. P O Box 14042, st Petersburg. Florida 33733 813-866 5151
_./ _
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Mr. Robert W. Reid Page Two May 22, 1980 Question 11.3:
...... Identify and describe those plant modifications to be made to automatically display core exit temperature informa-tion on the main control board.
Provide yot;r schedule and major milestones for the completion of these modifications."
Response
See response to Question 11.2.
Question II.4:
...... Identify and describe those plant modifications to be made to assure that the I&C needs associated with anticipated manual safety actions can reliably be met.
Provide your schedule and major milestones for the completion of these modifications."
Response
See response to Question II.2.
Question 11.5:
...... Determine and evaluate the performance of the ICS when either (a) NNI-X, (b) NNI-Y, (c) ICS-X, or (d) ICS-Y DC power supplies are tripped......Dccument and provide this evaluation to the NRC.
Identify and describe any modifications that this evaluation indicates are desirable. Provide your schedule and major milestones for the completion of these modifications."
Response
-See response to Question 11.2.
Question II.7:
...... Identify and descirbe those plant modifications to be made to eliminate spurious rupture matrix actuation.
Provide your schedule and major milestones for completion of these modifications."
Response
We have addressed this question in our submittal of May 2, 1980,- (Refer Item 51).
Additional infonnation has been pro-vided in our letter of May 6,1980.
Further information to complete our response will be forthcoming as delineated in our submittal of May 2,1980,. and letter of May 6,1980.
Mr. Robert W. Reid
' Page Three May 22, 1980 Question II.8:
" Describe those provisions that exist or will be installed to protect the pressurizer heaters such that loss of NNI power (or control circuitry) will not result in damage to the heat-ers if the pressurizer level is very low, i.e., the heaters left in a dry condition."
Response
See response to Question II.2.
Question II.9:
"...... Identify and describe any modifications you propose to preclude or mitigate single or multiple NNI control system ac-tions.
Provide your schedule and major milestones for the completion of these modifications."
Response
See response to Question II.2.
Question II.10:
"The failure mode values of control signals (e.g., T, T h
RC flow, 0TSG S/U level and operate level setpoints, etc.) ave.
and their collective effect on the RCS/ESF should be considered.
Discuss your plans to address this problem."
Response
See response to Question 11.2.
Question II.11:
"In response to No. 4 of your March 12, 1980, submittal, the OTSG S/U level, the operate level, and the steam pressure fail positions are identified as having no effect on control of
-plant.. During the Crystal River 3 February 26,1980, event, each one of those fail positions contributed to upset the plant control. Explain the discrepancy."
Response
See response to Question 11.2.
Mr. Robert W. Reid Page Four May 22, 1980 Should you have any questions concerning this subject, please contact this office.
Very truly yours, FLORIDA POWER CORPORATION 9h %"
P. Y. Baynard Manager Nuclear Support Services Attachment PYBhlcT0201
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STATE OF FLORIDA COUNTY OF PINELLAS P. Y. Baynard states that she is the Manager, Nuclear Support Services Department of Florida Power Corporation; that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Com-mission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of her knowledge, information and belief.
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afnard Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 22nd day of May,1980.
J Notary Public Notary Public, State of Florida at Large, My Commission Expires: August 8, 1983 PYBhlcT02D1 L.