ML20080P293
| ML20080P293 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/14/1984 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | John Miller Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR LIC-84-039, LIC-84-39, NUDOCS 8402220499 | |
| Download: ML20080P293 (4) | |
Text
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Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 402/536-4000 February 14, 1984 LIC-84-039 Mr. James R.
Miller, Chief U.
S.
Nuclear Regula tory Commission Office of Nuclear Reactor Regulation Division of Licensing Operating Reactors Branch No. 3 Washington, D.C.
20555
References:
(1)
Docket No. 50-285 (2)
Response to Section 2.2-2.4 of Enclosure 3, NUREG-0612 Transmitted to NRC by District Letter, Jones to Eisenhut (LIC-82-033), Dated January 21, 1982 (3)
Technical Evaluation Report Prepared by Franklin Research Center and Transmitted to District by NRC Letter, Clark to Jones, Dated May 17, 1983
Dear Mr. Miller:
Fort Calhoun Station Control of Heavy Loads The purpose of this letter is to formally submit changes to the scheduled completion dates for items identified as " Proposed Corrective Actions" on page 33 of Reference (2).
These commitment changes were discussed on January 16, 1984 and January 26, 1984 with Mr.
E.
G.
Tourigny of your staff.
In accordance with agree-l ments made during that telephone conversation, the new schedule and justification is as follows:
Proposed Corrective Action #1 A procedure will be written to provide an alternate path for shut-l down cooling water in the event of a load drop in the area bounded by Columns 10 and 11 and the biological shield wall in the contain-ment.
This procedure will permit the use of the polar crane in the area.
Commitment date:
January 21, 1984.
Present Status, Justification, and New Schedule This procedure is not yet finalized.
The present Plant Review Committee approved load handling procedure for the polar crane (OI-HE-1), however, prohibits load handling in the area bounded by Columns 10 and 11 and the biological shield wall.
This re-striction will not be removed until the proposed procedure for 8402220499 840214 PDR ADOCK 05000284 P
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1 Mr. James R. Miller LIC-84-039 Page Two providing an alternate path is approved by the Plant Reriew Com-mittee.
Further, the Fort Calhoun Technical Specifications pro-hibit use'of the polar crane for transporting of loads over the reactor' coolant system if the temperature of the coolant or steam in the-pressurizer exceeds 225*F.
Because of the restrictions currently in place and because the polar' crane will not be used until the refueling outage, the Distriut seeks an extension on this item until March 2, 1984.
Proposed Corrective Action #2 A procedure will be written to prevent the loss of the raw water pumps due to a load drop accident destroying the power supply
. ' cables.
The procedure will:
(a).
Prohibit loads from being carried over the area above the
~-24 cable tray supplying power to all four raw water pumps, and/or
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il '(b)
Outline _ emergency repair procedures to connect the fire pump
, ~
discharge idto_the raw water header to provide component cobling during the repair of the raw water power cables.
[
Commitment date:'" January 21, 1984.
g,,, Pros $nt St5tus, Justification, and New Schedule n-Loss of all raw water pumps is presently covered by an emergency procedure (EP-22),'" Loss of Raw Water".
This procedure requires a z' ' reactor. trip and' reactor coolant system cooldown to 300*F at the S.
maximum' allowable rate.
The. procedure also calls for putting pump AC-16^in service to'eupply wat'er to the demineralized water plant for' steam generator makeup. -This emergency procedure is now being revised.toealso allow the'use,of fire pumps to supply cooling water t'o the' raw water system (via temporary hose connections).
The-r'evised procedure is presently being evaluated by the Plant Review / Committee and is expected to be approved by February 10,
.;.1984. ;
_This~ delay will not have any effect on the plant safety because:
(1) EP-22 adequately addresse' ' loss of raw water pumps and (2) raw s
u water pumps'aro not. required for hot shutdown.
f
^ Pro [ooed lorrective' Action #3
'~The desisn"of Ihe access door to the reactor vessel cavity at ele-
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- ' vation 97,6'6" will be reviewed.
This design evaluation will en-sure that the door can withstand hydrostatic pressure of the flood-sed cavity after a, post 01ated loadLdrop shears off all nozzles of
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Mr. James R. Miller LIC-84-039 Page Three the reactor vessel and the vessel falls into the cavity.
If the door design is found to be deficient, appropriate steps will be taken to ensure the reactor core remains covered with coolant.
Commitment date:
End of the 1984 refueling outage.
Present Status, Justification, and New Schedule At the time this commitment was made, the District believed that a simple structural modification to the reactor vessel cavity door would be all that was required.
During the 1983 refueling outage, as-built conditions of the reactor vessel cavity door were check-ed.
Based upon this, the District realized that in addition to the door modification, the ventilation duct which penetrates through this F3or also needed strengthening.
Because of very high radiation levels in this area, the field measurements could not be completed.
The District plans to complete the measurements during the 1984 refueling outage and install this modification during the 1985 outage.
The District did not inform the Commission about this delay earli-er because installation and design difficulties in completing this modification have been very recently identified.
The District be-lieves.that because of very high radiation levels in this area and additional equipment and design effort required for sealing the ventilation duct opening, the full modification cannot be com-pleted during one outage.
It is the District's belief that an 18-month delay will not have any significant effect on plant safety because of the very low probability of a reactor vessel head drop accident.
This proba-bility is considered very low because of the very infrequent i
handling of the reactor vessel head (3 times during 18 months).
Further, this load handling is done by using the polar crane which is designed to meet applicable industry standards [page 23 of Re-ference (3)].
Further, all load handling within the containment i
building is governed by written plant procedures.
Personnel using l
this crane have been adequately trained.
The District's new pro-posed completion date is the end of the 1985 refueling outage.
Proposed Corrective Action #4 l
The geared rotary limit switches will be wired for the upper limit on the main hook and the auxiliary hook of the containment polar crane.
This will provide redundant limit switches and prevent a 2-blocking accident.
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g Mr. James R. Miller.
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" L IC;-84-039 Page Four-a 1
S ta tus ',
l' This was c~ompleted during-the 1983 refueling outage.
Since ely, f
W. C. J c nes Divisic.g Manager Product' ion Operations WCJ/DJM:jmm cc:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D.C.
20036 Mr.
E. G. Tourigny,-Project Manager Mr. L.
A. Yandell, Senior Resident Inspector L
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