ML19331C512
| ML19331C512 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 07/28/1980 |
| From: | Fiorelli G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Gerstner W ILLINOIS POWER CO. |
| References | |
| NUDOCS 8008180599 | |
| Download: ML19331C512 (1) | |
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GLEN ELLYN, ILLINOl$ 601M BUL 2 e syyz Docket No. 50-461 Illinois Power Company ATTN:
Mr. W. C. Gerstner Executive Vice President 500 South 27th Street Decatur, IL 62525 Gentlemen:
Thank you for your final report dated July 18, 1980, pursuant to 10 CFR 50.55(e) regarding Control Rod Drive (CRD) penetrations through the drywell wall. We will complete our review of this matter during a future inspection.
Your cooperation with us is appreciated.
Sincerely, f g [ ' ^ ^f?Q' Gaston Fiore111, Chief Reactor Construction and Engineering Support Branch cc: Director, RCI/II Director, AEQD Chief, OEB/MPA IE Files cc w/1tr dtd 7/18/80:
Central Files PDR Local PDR NSIC Mr. Dean Hansell, Office of Assistant Attorney General Mr. Gary N. Wright, Chief Director of Nuclear Safety 00Sh$0Af s
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O~llDELL U-0157 Ll4-80(07-18)-9 ILLINDIS POWER COMPANY 500 SOUTH 27TH STREET. DECATUR. ILLINOIS 62525 July 18, 1980' Mr. James G. Keppler Director, Region III Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
Reference:
IP Letter 6/24/80, U-0150, L. J. Koch to J. G.
- Keppler, NRC Region III; Subject Matter:
Extension of Time for 10CFR50.55(e) Final Report Clinton Power Station Unit 1 Docket No. 50-461 Construction Permit No. CPPR-137 On May 27, 1980, Illinois Power Company verbally notified Mr. H. M.
Wescott, NRC Region III, of a potential reportable deficiency per 10CFR50.55(e) associated with the Control Rod Drive (CRD) penetra-tions through the drywell wall.
By the referenced letter, we informed you on Jtne 24 by interim report that our investigation of the deficiency was not completed for final reporting.
We have since concluded our investigation as summarized in theWe attached report and corrective action is currently underway.
did not undertake a detailed investigation of the "reportability" of the deficiency as defined by 10CFR50.55(e) because:
(1) it would entail an extremely large number of analyses to cover all of the possible operating modes to be considered, and (2) the recommended corrective action was taken.
As the attached report indicates, CRD tube failure during normal power operation would not prevent normal control rod actuation and reactor shutdown, but under certain part load and reduced reactor pressure condi-tions the drive system may not function properly.
It would require expensive, time consuming analyses to evaluate each pos-sible operating scenario to determine if a reportable situation 4UL2 31980 p096 gco79Tso@6 m
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U-0157 Ll4-80(07-18)-9 2-We have concluded that it is more realistic to could result.
assume that a reportable set of circumstances could arise and to report the deficiency in accordance with the intent of 10CFR50.55(e?
All of the CRD tubes which had been adversely affected by the weld-ing are being replaced as corrective action for the deficiency.
the information provided in the attached final report We trust that is sufficient for your analysis and evaluation of the deficiency and the corrective action.
Sincerely, D
L. J. Koch Vice President LJK:nh attach.Director, Office of I&E, NRC, Washington, D.C.
cc:
H. H. Livermore, NRC Resident Inspector m
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a Attachmsnt to: U-0157 Y-5460 Ll4-80(07-10)-9 Clinton Power Station 10CFR50.55(e) Deficiency Report Control Rod Drive Penetrations Deficiency (10CFR50. 55(e) (1) (iii))
As a result of a cleanliness inspection of a control rod drive (CRD) penetration of the drywell (Penetration IMD-37), an area on the internal surface of Control Rod Drive Tube 2-B was found to have an area which appeared to have melted during the fillet welding pro-cess of the tube to the outside drywell plate.
Since the pipe (a purge was not a internals were not purged with an inert gas the melted area had a burnt sugar appearance (oxidized).
requirement),
Subsequently, all tubes in the CRD penetrations were inspected.
A total of fif teen insert tubes (1 1/4-inch Schedule 80 pipe, ASME SA-312, TP-316L) and twenty-three withdrawal tubes (1-inch Schedule 30 pipe, ASME SA-212, TP-316L) were identified by this investigation to have surface irregularities such as burn through and suck back on the internal surfaces of the tubes at the weld area connecting the tubes to the drywell penetration plate (1/2-inch carbon steel),
and several more tubes were found to have evidence of oxidation on the internal surfaces of the tubes in the weld area.
Corrective Action Nonconformance Report (NCR) No. 3041 was initiated and forwarded to Sargent & Lundy for dispositioning the deficiency.
The corrective action was to:
Remove all tubes identified as having burn through/ suck back a.
conditions utilizing Sargent & Lundy approved Job Instruction P-017.
Clean and polish the remaining tubes utilizing a Sargent & Lundy b.
Per the Job Instruction, the approved Job Instruction P-018.
tubes were to be cleaned and polished and reinspected after thirty days.
If a sugaring or rust-like condition reappeared after cleaning, a nonconformance report was to be initiated, Replace unacceptable CRD tubes in accordance with Sargent c.
Lundy approved design.
This corrective action is currently being implemented.
As of June 27, 1980, thirty-five (35) CRD cubes had been removed and the installation of the replacement tubes had commenced.
The NRC resident inspector, H. H. Livermore, has performed surveillance activities on this corrective action.
To prevent recurrence of these problems in the installation of replace-ment tubes, the following actions will be taken (per Job Instruction M-009, Rev. 0):
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The inside of the tubes will be purged with inert gas (i.e., argon) a.
during welding operations.
b.
During the welding operations, the welding variables (i.e., heat input) will be conitored.
Saferv I=clications The effect of the burn through and suck back on the integrity of the tubes over the life of the plant is very difficult to determine.
Since the tubes are stainless steel, it is believed that any tube failures would be in the form of cracks rather than a sudden rup:ure of the tube, and that all of the tubes would not fail si=ultaneously.
The reactor is designed so that the core syste= can be =ade sub-critical at any time or at any core condition with the highes: verth control rod fully withdrawn.
Thus, any CRD tube could fail totally and the reactor could be brought to cold shutdown (the =ost li=iting condition wi:h regard to the shutdows criteria) conditions.
If the reactor pressure was greater than 600 psig and all of the CRD tubes identified to have burn through or suck back were to fail to: ally, all control rods would insert on a scram signal.
Reactor pressure would cause the control rods with failed tubes to insert.
Thus, the reactor could be brought to cold shutdown conditions.
Damaged withdrawal tubes would not prevent a scram at any operating condition.
If reactor pressure was less than 600 psig and more than one inset; tube failed, the potential exists tha: the reactor could not be brought to cold shutdown conditions without using the standby liquid control system.
If this action is undesirable, the reactor could be maintained in a stable condition at a pressure less than 600 psig and the corresponding saturated temperature.
If reactor pressure would rise above 600 psig, the remaining withdrawn control rods could be inserted with reactor pressure and cold
. shutdown achieved.
A detailed engineering analysis has not been performed to determine what, if any, combination of insert tube failures (of those fifteen insert tubes identified) and reactor conditions would not give a cold shutdown condition without the standby liquid control system.
This is due to the cost of the analysis,
.the fact that all the tubes which had been adversely affected by the welding are being replaced, and such an analysis would not serve a useful purpose.
l The corrective actions for the deficiency assure that no unsafe installation exists.
Because a detailed engineering analysis was not performed, the significance of the deficiency to the safety of the plant has not been quantified.
In summary, due to the items list above:
a.
Failure mode via cracks is expected and not total failure.
b.
Any one rod could be fully withdrawn and the reactor could be made subcritical at all conditions.
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3-Above 600 psig reactor pressure, all control rods can be inserted c.
even if the CRD tubes fail.
The small fraction of time the reactor is below 600 psig, the reactor could be made suberitical with the standby liquid control d.
system if necessary or reactor pressure could be increased to 600 psig and then all control rods could be inserted.
A stable condition below 600 psig could always be reached.
e.
The safety significance of the deficiency is considered minimal.
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