ML19323G103
| ML19323G103 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/23/1980 |
| From: | Parker W DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8005300396 | |
| Download: ML19323G103 (4) | |
Text
DUKE POWER COMPANY Powen Bunm NO 422 SocTn Gnuncu Sruzer, CHARI.oTTz, N. C. 28242 mu w o. am e n. a.
May 23, 1980 Vace PatsiotNT TEttpwont: AntA 704 Straw Pacouctione 373-4083 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. B. J. Youngblood, Chief Light Water Reactors Branch No. 1 Re: McGuire Nuclear Station Units 1 and 2 Docket Nos. 50-369, 50-370
Dear Mr. Denton:
Mr. Robert L. Baer's letter of April 25, 1980 identified several staff require-ments pertaining to steam generator modifications and control rod guide tube pin replacement and requested that Duke inform the staff how these requirements would be implemented on McGuire. The purpose of this letter is to provide our position on each of these matters.
Control Rod Guide Tube Pins The staff noted that Westinghouse reported on March 13, 1980 that the Inconel 750 control rod guide tube support pins in Westinghouse plants may be suscep-tible to stress corrosion cracking and should be replaced.
This same matter was reported to NRC-Region Il by Duke Power Company on March 25, 1980 under the provisions of 10CFR 50.55e for McGuire Nuclear Station.
This was followed by a written repart on April 24, 1980.
In this follow-up report it was noted that the pins in use at McGuire were similar to those in which cracks have been discovered.
As a result these pins will be replaced before fuel loading with new pins which have been heat treated at a higher temperature to minimize the susceptibility to stress corrosion cracking.
Secondary Water Chemistry Monitoring and Control Program The position stated in your letter with regard to hotwell/ condensate monitoring is not clear.
The following discussion should provide additional clarification of the McGuire secondary water chemistry program. As stated in your letter the y
discharge of the hotwell pump at McGuire is continuously monitored. This sample point would be expected to provide the first indication for relatively large condenser leaks.
Other parameters may be more indicative of very small leaks.
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Mr. Harold R. Denton, Diract<>r May 23, 1980 Page Two When any parameters suggest a potential condenser leak, the individual hotwell sections are opled to confirm that a leak actually exists and to determine which condant.r section is leaking.
Once the leaking section is identified and isolated the teaking tube is removed from service.
r Our experience with stainless steel tubed condensers in plants with condenser cooling water similar to McGuire's has been quite good.
Nonetheless, over the plant life leaks will occur and the time required to confirm and find the leak will be dependent on the leak rate and failure mode. At McGuire, a large leak can be confirmed and located quite rapidly whereas a smaller leak may not be detected for some time due to the high quality condenser cooling water. The necessity of finding and repairing condenser leaks is recognized, however, we are not aware of any technical basis for a time limit for operation with a condenser tube leak.
The critical parameters in this situation are final feed-water and blowdown chemistry.
Since the McGuire system is equipped with full-flow condensate polishers and the condenser cooling water is relatively low in total dissolved solids a small condenser leak would produce little, if any, perturbation in these parameters and therefore would have no significant effect on steam generator integrity.
Should these parameters approach or exceed specification limits, applicable corrective actions would be taken.
In summary, we ffel that the secondary chemistry program at McGuire, including condensate and hotwell sampling, is adequate and provides the necessary controls to mini-mize the potential for steam generator degradation.
Further, to require a limit of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> for shutdown following confirmation of a condenser tube leak is arbitrary and could lead to steam generator operation under conditions for which immediate shutdown is warranted.
Steam Generator Modifications The staff's requirements consisting of plugging all the Row 1 tubes and cutting inspection ports in the McGuire steam generators are conflicting and totally unwarranted.
The basis for these requirements are observed failures in a relatively small number of operating Westinghouse Model 51 steam generators.
This matter was discussed extensively between Westinghouse and the NRC Staff in a meeting held April 15, 1980 in Bethesda, Maryland.
Subsequently, WesEinghouse provided additional information in a letter to Mr. R. H. Vollmer, Director, Division of Engineering, (NS-TMA-2241 dated May 12, 1980).
This letter provided l
a description of the forthcoming examination and testing program which Westing-house has undertaken to determine the mechanism causing some of the recent tube failures.
This letter also requested that NRC delay issuance of additional letters to near-term operating license plants until the "U" bend investigation i
currently in progress provides a sound technical basis for actions related to l
Row I tubes.
I Several utilities with Model D steam generators including McGuire have received letters from the NRC requiring plugging of Row 1 tubes. There is evidence indicating that the tubing in the Model D steam generator is significantly different from the tubing in the Model 51 steam generators and should not be-l l
Mr. Harold R. Denton, Director May 23, 1980 Page Three included in any generic assessment of the Row 1 "U" bends until additional data are developed. The tube size parameters for the two model steam generators are listed below:
Min.
Max. %
Tube Dia.
Bend Rad.
R/D Wall Thick Ovality Model 51 0.875 2.188 2.5 0.050 10 f
Model D 0.750 2.250 3.0 0.043 6
t As can be.een, the tube diameter, wall thickness, and maximum ovality are less in the case of the Model D tubing than for the Model 51.
The ratio of the tube diameter to bend radius is greater for the Model D.
All of these factors are in the direction of minimizing difficulties in the tube bending process and support the contention that the 3/4" diameter Model D tubing represent a different population.
i The NRC has also indicated concern with respect to Row I tube deformation as a result of tube support plate denting and has suggested that the Row I tubes also should be plugged in anticipation of subsequent denting of the units.
Operational experience has shown that plants with fresh water cooling are much less likely to encounter subsequent denting than are plants on sea or brackish water sites.
In the case of McGuire, the unit not only has fresh water cooling but also full flow,-
powdered resin condensate polishers. The likelihood of denting in these units is considered low.
In addition, should denting occur, it will be evident by normal inservice eddy current inspection long before it has proceeded to a point where it would result in deformation of the upper tube support plate flow slots and sub-sequent deformation of the Row 1 "U" bends.
At present there exist no definitive information as to the cause of the "U" bend failures which have been observed. At this point it is not even clear that the 4
"U" bend failures observed at the various sites are related. Therefore, any j
corrective action would be based on preliminary, speculative failure mechanisms.
The current Westinghouse program to analyze the cause of the "U" bend problem is expected to provide significant results by August 15, 1980, when examination of "U" bend samples removed from an operating plant will have been completed. The I
results of this program will permit decisions to be made on the basis of appro-priate technical data.
For these reasons it is believed highly inappropriate to remove approximately 3% of the tubes from operation by plugging the Row 1 tubes in the Model D steam generators.
4 Summary To summarize the Duke position on each of the identified requirements that are i
discussed in this letter Duke will:
1.
Replace the control rod guide tube support pins on Units 1 and 2 prior to-i fuel loading.
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Mr. Harold R. Denton, Director l
May 23, 1980 Page Four 2.
Utilize the existing secondary water chemistry monitoring and control program.
3.
Make no changes to the McGuire steam generators pending evaluation of the outcome of the Westinghouse steam generator tube testing program.
We would be happy to meet with you and your staff to discuss any of these matters further.
Very truly yours, 1
l h 4 lb <
WilliamO. Parker,JM CAC:scs
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