ML20196H332

Provides Further Basis for Fatigue Crack Growth Evaluation Supporting 880223 Request for Relief from Preservice Insp Requirements of ASME Section Xi.Final Rept on Analysis to Be Submitted 30 Days After Issuance of Full Power OL

ML20196H332
Person / Time
Site:	Braidwood
Issue date:	03/04/1988
From:	Ainger K COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
4316K, NUDOCS 8803100188
Download: ML20196H332 (2)

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) One Fast Nabonal Plara. Cficago, Ilkr6s O Address Reply to: Poet Offce Box 76T~

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\ March 4, 1988  !

) Mr. Thomas E. Murley, Director >

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC. 20555 Attn Document CoM col Desk '

Subject:

Braidwood Station Unit 2 Preservice Inspection (PSI) Program NRC Docket No. 50-457

Reference:

(a) February 23, 1988 letter from S.C. Hunsader to T.E. Nurley

Dear Mr. Murley ,

1 J This is to provide additional information regarding the basis for a '

relief request from the preservice inspection requirements of ASME Section XI. The relief request, contained in reference (a), pertains to an indication in the Braidwood Unit 2 loop 1 elbow to loop stop valve weld  ;

region. This relief request was supported by a fatigue crack growth .

1 evaluation documented in an interim report included in reference (a). (

T l At the request of the NRC staff, further explanation for the bases

.l of the fatigue crack growth evaluation of the weld indication in the interim I i report is provided in the attached addendum. The interim report contains a fatigue crack growth analysis based on the bounding conditions for the 4

reactor coolant piping system. The final report will be submitted with the fatigue crack growth analysis performed at the flaw-specific location and i will contain more details on the analytical methodology than pro"ided in the

! j 4 interim report. The final report will be submitted to the NRC within 30 t 9 days after issuance of the operating license authorizing full power ,

l operation for Braidwood Unit 2. f i

, Please direct any further questions regarding this matter to this offico.  :

l 4 Very truly yours,  !

i

. j K. A. Ainger  :

)

Nuclear Licensing Administrator

• i /klj att, cc: Braidwood Resident Inspector s WRC Region III Office i 4316K ,

t g 8003100188 980304 k PDR ADOCK 05000457 i 0 DCD i

o-Addendum Fatique Evaluation Bases in the Interim Report The fatigue crack growth evaluation provided in the interim report was for a typical reactor vessel inlet nozzle to pipe weld, instead of the actual location of the indication. This crack growth evaluation includes bounding pipe loadings, residual stresses, and the design thermal transients. The inlet nozzle to pipe weld is the reference fatigue crack growth location for the primary coolant system used in the Primary Loop Leak Before Break evaluations performed for the Braidwood Units. This location will give higher fatigue crack growth than the location of the indication in Braidwood Unit 2, because it is a terminal end, while the actual indication is not located at a terminal end weld.

Final Fatique Evaluation The fatigue crack growth evaluation to be performed and submitted in April will use the thermal loadings at the actual location, along with the appropriate piping loads and the residual stress pattern from the technical basis document for IWB 3640 (EPRI-NP-4690-SR).

4316K

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