ML20204C516
| ML20204C516 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/13/1978 |
| From: | Dail L DUKE POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811290227 | |
| Download: ML20204C516 (10) | |
Text
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DUKE POWER COMPANY
8 8 ' '828' GENERAL OFFICES
' ' ' ' ' " " ' ' s# a.4N 422 SOUTH CHURCH STREET l
CHARLOTTE. N. C. 28242 i
November 13, 1978 i
i Director of Nuclear Reactor Regulation Attn:
S. A. Varga, Chief Light Water Reactors, Branch 4 U. S. Nuclear Regulatory Comission Washington, DC 20555 Re: Catawba Nuclear Station Steam Generator Compartments Analyses Duke Files: CN-1147.00 and CN-1412.05
Dear Mr. Varga:
As requested in your Decembe.r 2,1977 letter and in subsequent telephone discussions with Messrs. Kane, Kiessel, Bosnak and Cherny, Duke submitted information and analyses concerning Catawba Nuclear Station steam generator 13, 1978. Your letter of October 6,1978 indicated compartments on March that while most of our submittal was. generally acceptable, four items of further information were required.
Enclosed are Duke's responses to the four questions of your October 6,1978 letter.
Very truly yours, 1
L. C. Dail Vice President Do 'm Engineeri.ng JEB/jmi Enc 1c3ures Qo(h i
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(Page 1 of 4)
RESF0NSES TO NRC QUESTIONS (Varga letter of 10-6-78) 1.
Your analysis of subcompartment pressures is identical to corresponding portions of the analysis used in McGuire.
However, the McGuire analysis also included a nodalization sensitivity study, and that plant's maximum calculated differential pressure for the steam generator enclosure sub-compartment design of 13.75 psi was determined from a two-node model.
The Catawba analysis refers to a nodalization study, but the steam generator enclosure peak differential pressure is shown as 12.5 psi (for the nine-node case).
Please explain this apparent discrepancy, or use 13.75 psi as the basis for the design of the steam generator enclosure.
It should be noted that on Catawba you have agreed to apply a 40%
margin to the maximum calculated pressures for u9 in subcompartment design. Therefore a design pressure differential of 1.4 X 13.75 = 19.25 psi would appear appropriate for the Catawba steam generator encNsures.
DUKE'S RESPONSE The steam generator enclosure subcompartment r.odalization accepted for use on Catawba Nuclear Station was the nine (9) node model. The model was described in our submittal to the USNRC transmitted by Mr.
W. H. Owen's letter dated March 13, 1978. This subcompartment nodalization scheme was chosen such that pressure gradients existing within each node are negligible.
Verification of this was accomplished by conducting a sensitivity study that included not only a two (2) and nine (9) node model, but also a one (1) node model.
Review of the two (2) node model results indicated that one of the nodes contained pressure gradients which resulted in unnecessarily conservative values for the computed peak pressures.
The nine (9) node model eliminated these nodal pressure gradients resulting in a more accurate calculation of the peak pressures.
The objective of our nodalization study was to arrive at the most accurately computed value for the maximum differen-tial pressures for design of the steam generator enclosures. This was accomplished with tne nine (9) node model which resulted in a maximum differential pressure of 12.5 psi.
Duke Power Company defines the minimum subcompartment design differ-ential pressure as the preliminarily calculated subcompartment differ-antial pressure increased by a factor of 1.4.
This approach is consistent with our commitment contained in Catawba PSAR Section 6.2 as well as SRP 6. 2.1. 2.
(Page 2 of 4) 4 RESPONSES TO NRC QUESTIONS (Varga letter of 10-6-78) 2.
To be consistent with our position regarding subcompartment analysis margins, the maximum asymmetric pressure acting on the steam generator should also have a 40% margin applied for use in design of the supports.
Please confirm your agreement with thic position.
DUKE'S RESPONSE. Page 8 of the TMD Code Section of the March 13, 1978 submittal tabulates the preliminary calculated subcompartment differ-ential pressures across the steam generator vessel.
These pressures are increased by a factor of 1.4 to establish the minimum subcompartment design differential pressures for computing steam generator support design loads.
4 1
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RESP 0NSES TO NRC QUESTIONS (Varga letter of 10-6-78) 3.
With regard to inservice' inspection for the main steam pipe welds enclosed by the guard pipe, you state that " inspection ports must be provided for these welds." We interpret this statement as a commitment to provide such inspection openings.
Please verify this r
interpretation.
DUKE'S RESPONSE There are five guardpipe-enclosed welds (welds 1-5 on Attachment 3) subject to volumetric ISI on each main steam line:
four of these five welds will be made accessible by inspection ports.
shows a typical inspection port arrangment. Weld No.1 can be made i
accessible, if necessary, by the removal of the nine-inch guard pipe i
extension. Weld No. 2 is totally inaccessible because of the restraint pads inside the guard pipe. This is shown on Attachment 3.
Since making Weld No. 2 accessible would require major modification to the guard pipe and restraint configuration'- causing possible degradation of the entire system - it is requested that this weld be exempted from volumetric examination. A separate Request for Relief in accordance with NRC guidelines will be submitted with the preservice examination program.
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RESPONSES TO NRC QUESTIONS (Varga letter of 10-6-78) 4.
Certain features of your inservice inspection plan are not considered acceptable at this time. An acceptable plan would include the following:
a.
Piping System Welds Other Than in the Containment Isolation Region You should provide a commitment to base preservice and inservice inspections on the Edition of ASME Code Section XI required by 10CFR Part 50.55a(g).
b.
High Energy Piping Systems Welds in the Containment Isolation Region Preservice and inservice inspections should be based on the augmented ISI requirements defined in SRP Section 6.6.
When gut a pipe is used in this region, you should provide sufficient access, by either inspection ports or removable guard pipe, to perform the required augmented inservice inspections.
DUKE'S RESPONSE
- a. The perservice examination will be performed in accordance with the ASME Boiler and Pressure Vessel Code,Section XI 1974 Edition through Summer 1975 Addenda. The inservice inspection program for the first forty months of the first ten-year interval will be in accordance with the Section XI Code in effect as required by 10CFR50.55a(g) to the extent practical.
Specific Request for Relief from Secticn XI requirements wi be submitted in accordance with NRC guidelines.
- b. Currently there are no welds, other than those described in question 3, subject to augmented inservice inspection as described in Standard Review Plan Section 6.6.
The Catawba Penetrations, which previously were subject to this examination, have been modified to allow the exemption specified in SRP Section 6.6, Paragraph II.8.d.
At' ichment 4 is the intended ISI program for the Catawba Unit 1 Main Steam System and is included for information only. Attachment 4 supersedes the Section XI Item No. 2.1 program included in the previous submittal.
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