ML20088A726
| ML20088A726 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/06/1984 |
| 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-84-094, LIC-84-94, NUDOCS 8404130019 | |
| Download: ML20088A726 (7) | |
Text
l Omaha Public Power District 1623 Harney Ornaha. Nebraska 68102 402/53E 4000 i
April 6,1984 LIC-84-094 Mr. James R. Miller, Chief U. S. Nuclear Regulatory 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 Secti, n 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)
District Letter, Jones to Miller (LIC-84-039) dated Feb rua ry 14, 1984.
Dear Mr. Miller:
Fort Calhoun Station Unit No.1 Control of Heavy Loads, Phase 2 In accordance with Reference 2 (page 33, item 1), the Omaha Public Power District made a commitment to develop vritten procedure to provide an alternate path for shutdown cooling water in the event of a load drop in the area bounded by Columns 10 and 11, and the biological shield wall, in containment. This procedure would then allow use of the polar crane in this area.
Reference (3) changed the completion date for this commitment so that additional calculations verifying the validity of the proposal could be performed. These calculations were completed and, as reported to Mr. E. G. Tourigny of your staff in a telephone call on February 29, 1984, revesled that the proposed solution is not acceptable. The District has further investigated the possible alternatives and provides the following discussion and revised commitment.
The commitment was initially made to meet the requirements of Enclosure 3, Section 2.4 of NUREG-0612, which addresses crane operation in the area where equipment for decay heat removal is located. The area of concern contains the Safety Injection and Charging headers which are required for shutdown cooling and boron injection.
On further evaluation it has been determined that because of the following reasons, use of the proposed alternate flow path is not the best alterna-tive for meeting the requirements of NUREG-0612 Enclosure -3, Section 2.4:
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P PDR I I 45.5124 Employmen h Equal opportunity
T Mr. James R. Miller April 6,1984 Page Two l
1.
The alternate flow path requires that the flow must be routed through a 1" line at a very high rate. This may result in excessive vibra-tions leading to failure of this line if used for an extended period.
4 2.
The flow rate provided is not adequate to remove the decay heat from the core until 45 days from reactor shutdown, and after 1/3 of the core has been removed. Thus, the current load restriction would need to be maintained until 45 days after reactor shutdown.
3.
In the event of a load drop the alternate path would have to be phys-
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1 ically realigned to become operational. Many of the valves requirir.g alignment are locked closed, and a time factor could become important to get the system operational.
l
. 4.
This procedure could only be used when the reactor head is removed from the core, and the water level is above 15 feet.
L Because of the above, we believe that this procedure will be unduly restric-tive and will not be practical. Therefore, we would like to revise this commitment and exercise other options provided in NUREG-0612., Subsection 2.4.2.b of NUREG-0612, provides three options to meet the established criteria. These options are:
1.
If separation and/or redundancy is provided between safety-related equipment, then the hazard is eliminated in the event of a load drop.
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- 2.
Where mechanical stops or. electrical interlocks are provided to pre-vent carrying loads over safe shutdown equipment, the hazard can be eliminated.
3.
Where technical specifications or administrative procedures are.used to eliminate the load hazard, discussion should be provided to ensure validity of the constraints.
In accordance with the existing plant procedures, the District is in complf-ance with requirements of Enclosure 3, Section 2.4 by meeting the require-ments outlined in subsection 2.4.2.b option 3 above.
.An adninistrative procedure (OI-HE-1) currently exists which restricts load
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handling in the area defined above. - This procedure has been in effect
- - since May 8,1981. This procedure administrative 1y prohibits loads from being carried in the hazard area.- This restriction can only ~be overridden for a limited period or for handling a specific load per a written proce -
dure with Plant Review Cammittee approval. This ensures that probability of a load drop in this area is extremely'small, which is consistent with
' the. requirements of NUREG-0612 Enclosure 3, Section 2.4.
i.
Attached please find a copy of'the revised pages to the District's Section r
2.2-2.4l response, which : incorporates the results of the above. discussion.-
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't Mr. James R. Miller April 6,1984 Page Three This' revised corrective action requires use of administrative controls to meet Section 2.4.2.b requirements. Further, these change pages incorporate the current status of Proposed Corrective Actions, Items 2, 3, and 4, as it was detailed in Reference 3.
Sincerely, NM W. C. Jones Divisijn Manager Production Operations WCJ/MDH/rh-N 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 4
ATTACHMENT Revised Pages to the District's January 21, 1982 Letter 9
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0 4.0 PROPOSED CORRECTIVE ACTIONS The following corrective actions are planned based upon the results of the design evaluation and analysis of Control of Heavy Loads at Nuclear Power Plants.
1.
The load handling operations by the polar crane shall continued to be governed by the existing procedure (01-HE-1). This pro-cedure restricts load handling in the area bounded by Column Lines 10 and 11 and the biological shield wall in the contain-1-
ment (Ref. GHDR #11405-A-5, Appendix B).
This restriction can only be removed for a limited period or for handling a specific load per a written PRC approved procedure. This procedure has been implemented since May 8,1981, as an interin corrective action to meet the requirements of HUREG-0612.
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 cable traf supplying power to all four raw water pumps, and/or b)
Outline emergency repair procedures to connect the fire pump discharge into the raw water headar to provide com-ponent cooling during the repair of the r6 iter pump power cables.
This item is complete; see the District's letter dated February s
I 14,1984 (LIC-84-039).
3.
The desipn of the access door to the reactor vessel cavity at EL. 976'-0" (Ref. GHDR Drawing No.11450-A-14, Appendix B) will be reviewed. This design evaluation will ensure that the door can withstand hydrostatic pressure of the flooded cavity after a postulated load drop shears off all the nozzles of the reac-tor 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.
4.
The Geared Rotary Limit Switches will be wired for the upper limit on the main book and the auxiliary hook of the contain-7 ment polar crane. This will provide redundant limit switches and prevent a two-blocking accident.
This item is complete; see the District's letter dated February 14,.1984 (L10-84-039).
h 5.0 -SCHEDULE
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Proposed Corrective Action, Item 3 will be completed by the end of b
the 1985 Refueling Outage as was detailed in the District's letter dated February 14,1984 (LIC-84-039).
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CRANE CONTATPhlENT PollA CRANE
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C0trr/dNHENT r-in HEAVY LOAD Rv NEAD s ra Fr Ric 120 T, i.0WER i.0An'
-lMPAC<T AFd.'A SEE FICURES IN APPENDIX B AND FIGUltE 11,
ill.DCK 5T (MOVED DURING REFUELTNG OUTAGE ONI.Y)
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ELEVATIOf4 SAFETY-REl ATED,.;
llAZARD EfidlNATION STATEMEN T.
/EOUIPMENT.
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llCV-327, llCV-329,11CV-MI,1d -333, AND IICV-348 TilESE VAINES Atin TIIE VAINES OUT-y,,
1dDE OF CONTAINilENT CAN. Ilt Cl.0 SED TO STOP DRAINING OF Tl!E SYSTEtt OR Tile REAC i
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CAVITY., ' TiiE TECil. SPECS. AI.T.OW 8 Il00RS FOR Tile SYSTEM To i1E INOPERAllt.E IF NOT
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IN REI"JELING. TIIEREFORE, Tile PLANT COULD WEI.D A III.IND FI.ANCE ON lilE NECESSARY PIPING' ENDS AND PLACE A PORTION OF Tile C001.DOWN SYSTEtt!!ACK IN OPERATION IN T 8 110U11S TIME I.IllIT.'..NO IIAZARD FOR CASE (I).
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1013' IF LOAD DROP SilEARS VAINE llCV-348,. SYSTD8 LOW POINT AT E. 1003', TIIE REACTOR CAVITY COUI.D DRAIN, WIT!! N0 OPEAATOR ACTION, TO Tile DOTTot! 0F TIIE Il0T l.EG rirE t
AT AN APPROX. E!.. OF 1005'.
THIS i/0UI,D LEAVE APPROX. 4' OF WATEli COVEI,tINil THE CORE.
IT S110UI.D 11E NorED, NOWEVE;1, TilAT Tile I.OWER CAVITY AREA WII.I. NOT DRAIN IlEl.0W 1013' DUE *10 Tile CONCRETE WAI.I. SEPAlt ATING Tile VESSEL. FROM TIIE I.0WE!! C Tile TECil SPECS AI.I.0W Tile SYSTFll TO 11E INOPERAlli.E FOR A ftAX11111tf 0F S llulfitS AND Tile Pl. ANT STAFF COIII.D REPAIR Tile IIREAK WITillN TilAT TI!!E PERIOD...NO IIA 7.AltH Full CASE (II).
- 9) 994'-0"
- (2) SAFETY-INJECr!ON (2)
The probability of a load drop in this area is extremely small since a
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AND CilARGING PIPING restriction has been placed on the Polar Crane prohibiting carrying loads in this area.
The restricted area falls between column lines 10 and 11 on the drawing attached, the containment wall, and steam genera tor RC-28 wall.
This procedure administratively prohibits loads from being carried in the hazard area.
This restriction can only be overridden for a limited period or for handling a specific load per a written Plant Review Commit-tee approved procedure.
This ensures that probability of a load drop in this area is extremely small, which s consistent with the requirements of NUREG-0612 EQUIPMENT REFERENCED TO THIS tl0TE IS THE ONLY SAFETY RELATFn F0l'IPMENT RE0 HIRED TO MEET NUREG-0612 CRITERIA WHICH IS LOCATED WITHIN THE LOAD DROP ENVELOPE.
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1.0 AD / DROP TABl_E - A-1 2 or 3
CRANE' CONTAINHENT POI.AR CRANE (CONT.)
. LOCATION CONTAINHENT l-lEAVY LOAD RV HEAI) 4 I.TFT R1C 1207. I.0WER I.0AD MLOCK IMPACT AREA SEE FICuRes IN APPENoix a ANu FICuRE n-1 ST SAF{E ELATED
~ ELEVATION IlAZARD ELIMINATION STATEMENT,.
(2) 994'-0"-
- (2) SAFFTY-IN. LECT 10N CONTItillED CilAHGING l'IPING
' Enclosure 3, Section 2.4.
This administrative procedure meets the require-ments of guideline 2.4.2.b(3).
The ad:riinistrative procedure has been in ef fect since May 8,1981.
No hazard, therefore, exists.
1 (1) 1013'-0"
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REACTOR VESSEL.
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'lllE VESSEL WOUI.ll TilEN DR0l' INTO lilE REACTOR VESSEL CAVITY.
TIIE CAVITY Wotll.D Fil.l. WITil WATER FROt! Ti!E ILROKEtt 1.00P N0ZZI.ES.
ADDITIONAI.
WATER WOUI.D llE SilPI'I.IED Fit 0tl Tile ACCllittil.ATOltS APID SIR'J TAtlK I F 1:r.f}lIlltr.n.
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