ML20059D374

Advises That Util Plans to Model Bvps,Unit 1 FW & Extraction Steam Sys Susceptible to Flow Accelerated Corrosion Not Modeled at Bvps,Unit 2,in Response to NRC Re Insp Rept 50-412/93-21

ML20059D374
Person / Time
Site:	Beaver Valley
Issue date:	12/30/1993
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9401070203
Download: ML20059D374 (1)

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_N B*ever Valley Power Station

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OB Sheppmgport, PA 15077-0004 l

December 30, 1993 JOHN D. SIEBER (412) 393-5255 Senior Vice President and Fax (412) 643-8069 Chef Nuclear Officer Nuclear Power Dmsson U.

S.

Nuclear Regulatory Commission Attn:

Document Control Desk Washington, DC 20555

Subject:

Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No._NPF-73 NRC Inspection Report 50-412/93-21 The NRC letter dated November 1,

1993, which transmitted Inspection Report 50-412/93-21, noted that Duquesne Light Company l

(DLC) was evaluating predictive methods for selecting components for i

erosion / corrosion monitoring at Beaver Valley Unit No. 1 (BV-1) and i

Unit No.

2 (BV-2) and requested our plan and schedule for implementing an expanded program.

l DLC has completed the modeling of the BV-2 Feedwater and Extraction Steam systems using the EPRI endorsed "CHECMATE" computer program.

Before the BV-1 tenth refueling outage, DLC plans'to model

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the BV-1 systems susceptible to flow accelerated corrosion (FAC) that have not.yet been-modeled at BV-2.

A: review to verify design.

similarities between the BV-1 and BV-2 systems will then be performed to ensure that-the conclusions determined for the modeled system at one unit are also valid for the corresponding system at the other unit.

The FAC examination selections for the BV-1 tenth refueling outage and the BV-2 fifth refueling outage will be based on the "CHECMATE" modeling results.

The data obtained from these FAC-j examinations will be integrated into the models to improve wear rate predictions.

A review will be conducted at that time to determine if i

the modeling effort needs to be expanded.

The scope of future FAC examinations will be based on both observed wear rates and model results.

If you have any questions concerning this

response, please

• ^ntect Mr. Greg Kammerdeiner at (412) 393-5677.

Sincerely, 050052 h pnf k J. D.

Sieber 7'

cc:

Mr.

L. W.

Rossbach, Sr. Resident Inspector Mr. T. T.

Martin, NRC Region I Administrator fhf Mr. G.

E.

Edison, Project Manager THE MMR I

Mr. J.

P.

Durr, Chief, Engineering Branch Division of Reactor [ Safety, Region I ggf 0' 9401070203 931230 PDR ADDCK 05000334 g,

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PDR k,

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