ML20211C807
| ML20211C807 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 09/30/1986 |
| From: | Ainger K COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2171K, NUDOCS 8610220026 | |
| Download: ML20211C807 (4) | |
Text
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Commonwrith Edson
_.,, 72 West Adams Street, Chicago, Illinois
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/ Address Reply to: Post O!!iceTo'x767
'U Chicago, Illinois 60690 - 0767 t
September 30, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Station Unit 2 Preservice Inspection NRC Docket No. 50-455 Reference (a): May 21, 1986 letter from K. A. Ainger to H. R. Denton
Dear Mr. Denton:
Reference (a) provided a report on the final disposition of ultrasonic indications identified in the Byron Unit 2 steam generators and pressurizer during.the Preservice Inspection. During discussions with the NRC staff on September 17, 1986, additional information was requested concerning 12 unacceptable indications in the steam generators which are described in reference (a). Enclosed are Commonwealth Edison's responses to six questions raised by the NRC staff.
One signed original and fifteen copies of this letter and enclosure are provided for NRC review.
Please direct any questions regarding this matter to this office.
Very truly yours, K. A. Ainger Nuclear Licensing Administrator im Enclosure cc: Byron Resident Inspector
@oO I
I 8610220026 860930
{DR ADOCK 05000455 2171K PDR
Attachment Additional Nuclear Regulatory Commission Concerns Regarding Byron. Unit 2 Steam Generator Weld Indications
Reference:
Letter from A.D. Miosi to H.R.
Denton dated February 28, 1986 The following questions were asked by the Nuclear Regulatory Commission (B. Elliot) in a conference call with Commonwealth Edison (CECO) on September 17, 1986.
Ceco's responses to these questions follow:
1.
Question:
Are the ultrasonic characteristics of the eleven (11) rejectable indications remaining in the steam generators similar to
.the ultrasonic characteristics of the indications removed by grinding or core sampling?
l CECO Response:
The ultrasonic examination results of the rejectable indications remaining in the Byron Unit 2 steam generators are similar to the ultrasonic examination results of the indications removed by grinding or core sampling.
The ultrasonic length j
dimensions of the removed indications range from 0.35 to 1.0 inches and their ultrasonic through-wall dimensions range from 0.24 to 0.53 inches.
All except one rejectable indication have length and through-wall. dimensions that fall within'these ranges or are smaller.
The exception is the rejectable indication in steam i
generator 2096, weld SGC-05 with a through-wall dimension of 0.60 inches.
However, this indication is subsurface with an a/t value of l
only 8.8%.
This a/t value is toward the lower range of the through-wall extent estimates for the removed indications (6.0%-12.4%).
CECO has concluded that the eleven rejectable indications have the same ultrasonic characteristics as the indications which were removed.
2.
Question:
Are the inner diameter surfaces of the welds containing the three (3) rejectable surface indications accessible?
l CECO Response:
The inner diameter (I.D.) surfaces of the welds containing the three (3) rejectable surface indications are not accessible.
Two (2) of the three (3) surface indications are located in stub barrel-to-tubesheet welds.
Since this weld is located just above the tubesheet, the steam generator tube bundle and tube bundle wrapper prevent access to the weld I.D.
surface.
Consequently, these indications cannot be removed from the I.D.
surface by grinding without removing major components of the vessels.
The third t
rejectable surface indication is located near the I.D.
surface of a l
main feedwater nozzle-to-shell weld.
Access to the I.D.
surface of this weld is prevented by the preheater section of the steam I
i generator.
Again, removal of the indication by grinding and welding involves unusual hardship.
I 1
l 1 3.
Question:
Why were the 11 rejectable indications not removed by coring similarly to the removal of the weld samples?
CECO Response:
CECO has concluded that the alternative of removing the indications by coring is unnecessary base on the following:
All of the indications have the same ultrasonic characteristics as those that were explored by core sampling and mechanical removal by grinding.
The ultrasonic information used in the analyses of the remaining indications conservatively oversize these reflectors.
Six of the indications are located at structural discontinuities such as the transition cone weld and the feedwater nozzle where repair by creation of a hand hole is impractical.
One of the indications is located in the stub barrel-to-shell weld (SGC-03) where there is a change in the outside diameter of the shell.
Creation of a hand hole in this location may require a weld pad buildup to compensate for this taper.
The other five indications are located in the stub barrel to tubesheet weld (SGC-02).
Two of these indications are located closely enough that an oversized hand hole (greater than 3 inch diameter) would be required to ensure removal.
This is likely to require a weld pad buildup to accommodate a hand hole cover plate.
There were originally seven indications in stub barrel to tubesheet welds (Welds SGC-02).
Two of them were the locations where exploratory core samples have been extracted and metallurgically examined.
The results showed the indications resulted from small slag inclusions and were oversized by ultrasonics.
There is no compensating additional information to be gained by more core sample exploration.
Two of the original five locations where this was clearly possible without subsequent weld repair (isolated indications in Weld SGC-02) have already been explored.
The additional of more hand holes will create extra opportunities for secondary system leakage.
l 4.
Question:
Why were thermal stresses due to thermal stratification not considered in the fracture mechanics analysis of the feedwater nozzle weld configuration?
l l
l l
t l
I
. CECO Response:
It is not necessary to include such thermal stresses in the feedwater nozzle analysis because the design of the Byron steam generators precludes thermal stratification at the main feedwater nozzle.
Low temperature feedwater will be provided at low flows to the steam generator only through the auxiliary feedwater nozzle.
The feedwater will be sufficiently heated before entering the vessel through the main nozzle.
Therefore, the layering of hot and cold feedwater will not occur in the main nozzle.
The design temperature' transients for main feedwater nozzle have been accounted for in the fracture mechanics analysis.
5.
Question:
Is the inner diameter surface of the steam generator transition cone upper weld (SGC-06) covered with water during normal plant operation?
Ceco Response:
Both Byron Unit 2 and Braidwood Unit 1 will operate with constant water level programs for the steam generators.
During plant operation ranging from 0% to 100% reactor power, the steam generator water level will be above the transition cone upper weld.
Therefore, the inside diameter surface of the weld will be covered with water at all power levels.
(Note that CECO's response to this question for Braidwood Station in the referenced letter is incorrect.
The above response is correct for Braidwood.)
6.
Question:
What is the copper content in the feedwater system at Byron Unit 2?
CECO Response:
Components of the feedwater system at Byron Unit 2 contain no significant amounts of copper or copper alloys.
Tubing of all feedwater system heat exchangers and main condensers are made of Type 304 stainless steel.
Therefore, copper corrosion in the Byron feedwater system will not occur.
8503b