ML20138C715

Forwards B&W Evaluation of 850906 Pressurization of Once-Through Steam Generator 1.1.Event Had No Adverse Structural Effect & Steam Generator Remains Acceptable for Continued Operation at Full Power Conditions

ML20138C715
Person / Time
Site:	Davis Besse
Issue date:	10/18/1985
From:	Williams J TOLEDO EDISON CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
1198, TAC-60062, NUDOCS 8510230013
Download: ML20138C715 (22)

Text

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l Docket No. 50-346 License No. NPF-3 TntEDO

%mm EDISON Serial No. 1198 JOE WILUAMS. Jn October 18, 1985 s~ **== e<

(4'9) P49 2300 l41% 243 52?3 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz Operating Reactor Branch No. 4 Division of Licensing United States Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Stolz:

On September 6,1985, during special testing of the Davis-Besse Nuclear Power Station Unit No. I cuxiliary feedwater pump turbine (AFPT) 1-1, the secondary side of the once-through steam generator (OTSG) 1-1 was subjected to a maximum pressure of 1058.5 psig. This cold pressurization was discussed and reported to the NRC in Licensee Event Report 85-017 submitted October 4, 1985.

Subsequent to this event, an engineering evaluation was performed as required by Section 3.7.2.1.b of the Davis-Besse Technical Specifications.

This evaluation included an analysis of the potential growth of existing flaws in the secondary side shell and main steam nozzle regions by Babcock and Wilcox (B&W). Per Mr. A. W. DeAgazio's-recent request, copies of these B&W analyses are enclosed (Attachments 1 and 2).

t In addition, Toledo Edison management determined that it was prudent to review the existing analysis of structural loadings associated with water filled main steam lines. Bechtel Power Corporation reviewed the l

structural loadings associated with water filled main stean lines

, performed as part of the original design analyses. Bechtel has confirmed the structural adequacy of the main steam lines and supports to withstand the effects of being filled with water.

Based upon review of the transient and completion of stress and fracture mechanics analysis, Toledo Edison conducted a 10CFR50.59 review. This I review concludes that the September 6, 1985 pressurization of the OTSG 1-1 I had no adverse structural effects and that the steam generator remains acceptable for continued operation at full power conditions. Toledo Edison has concluded that this is not an unreviewed safety question.

Very truly yours, i

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• H "' REVIEW OF SECTION XI EXAMINATION DATA ON TED, SEPTEMBER 09, 1985 DAVIS BESSE, STEAM GENERATOR l-1 A review of our records shows that the Section XI baseline was performed in accordance with the 1970 Edition of Section XI with Addenda through the winter of 1971. As such, the baseline code only required ultrasonic examination of the class 1 weldments.

Subsequently, the ISI program was updated to the 1977 Edition of Section XI with Addenda through the summer of 1978. This edition of the code required examinations of certain Class 2 (secondary side) weldments.

During the second outage (summer of 1982) of Davis Besse, two of the Class 2 welds on steam generator 1-1 were ultrasonically examined. The Davis Besse ISI examination program identifies these welds as: (1) figures number C1.010.001, shell to shell, MK1 to MK2, and (2) figure number C1.030.001, upper tubesheet to shell, MK51 to MKl. The examinations included the full circumference of each weld.

The weld identified as figure number C1.010.001 was determined to have several laminar flaws all of which are of acceptable size in The weld identified as figure accordance with standard IWB-3511-2.

number C1.030.001 had only geometric (non-relevant) reflectors which represents an acceptable condition. These are the only examinations that have been performed on the secondary side of steam generator 1-1 to date.

Based on the examination results reported, the postulated 1/4T flaw size used in your analysis of the over pressurization of the generator is considered conservative in that it bounds all of the data reported during the ISI.

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Toledo Edison Company has informed Esw that in the process of testing the secondary side of the OTSG on 9-6-85100 that 0 F. they Thic inadvertently achieved a pressure of 1030 psig at pressure and temperature are less severe than the code hydrotest conditions, however, the steam line was filled with water. De-canoe of the additional weight effects, the caleclations here-in were performed to justify the condition on the nozzle.

a.que coa. n auir...nts The stea= outlet nozzles were originally analized as part of a Class 1 vessel (Per Ref.1) . The applicable Code was AS?m,1968 Edition Mith Addenda thru Sammac 1968 (Ref.1) .

From Para. N-713.2, the applicable stress limits are P, < 0.9 S y ,

P3+Pg < 1.35 Sy More recently, the 1980 Edition of the code specifies stress requirements for pneumatic test to bei 2,< 0.9 Sy Pg+P3 < 1.35 Sy when P,< 0.67 Sy

< 2.15 57 - 1.2 P, when 0.67 Sy<Pm < 0.9 Sy r.a n M a g e the loada on the nozzle due to the added weight of the water have '

been provided (Ref. 3) and are Ma B norrla F, P b

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