ML19332C715
| ML19332C715 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/15/1989 |
| From: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8911280462 | |
| Download: ML19332C715 (3) | |
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.i BALTIMORE GAS AND-ELECTRIC CHARLES CENTER
- P. O. BOX 1476
- BALTIMORE, MARYLAND 21203 GEORGE C. CREEL Vict PaassotNT Nuctran C=cnov.
aoo no un November 15,1989 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs. Nuclear. Power Plant Unit No.1; Docket No. 50-317 Relief For Main Steam Pioinn Flaw at Calvert Cliffs Unit 1
REFERENCES:
(a) Letter from Mr. A. R. Thornton (BG&E) to Mr. Ashok C. Thadani (NRC), dated December 16, 1986, Request for Approval of Main Steam Piping Evaluation per ASME Section XI (b) Letter from Mr. J. A. Tiernan (BG&E) to Document Control Desk, dated September 30, 1987, Main Steam Piping Flaw at Calvert Cliffs l
Unit 1 (c) Letter from Mr. W. J. Lippold (BG&E) to Document Control Desk, dated May 4,1988, Main Steam Thin Wall (d) Letter from Mr. S. R. Cowne (BG&E) to Document Control Desk, dated June 6,1988, Main Steam Thin Wall 1
Gentit. men:
i We are seeking final resolution of the issue of the flaw in the main steam piping of our Unit 1 Number 12 Steam Generator. On August 17, 1989 one of our senior mechanical i
design engineers-met with our NRR Project Manager to discuss remaining concerns. They determined that more information. is required concerning our fracture mechanics analysis, if it is to be credited. Also, the question should be addressed as' to whether a Steam Generator Tube Rupture event would result in our steam piping being subjected to Reactor Coolant System pressure prior to that pressure being lowered below 1000, psig, the design pressure of the main steam piping. These items are discussed below.
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Document C:ntr:1 Desk November 15,1989 Page 2 ANALYSES OF Tile FLAW BACKGROUND Reference (a) submitted our initial assessment of the Main Steam Thin Wall issue. The submittal reported that all primary longitudinal stress levels are within USAS B31,1-1967 (B31.1) Code allowables for each load - combination. More specifically, all thermal, deadweight, longitudinal pressure, and seismic - loadings are acceptable. The submittal reported that, per the requirements of B31.1, paragraph 102.2.4(2), the primary transient hoop stress limits are met for all accident conditions iisted in our Updated Final Safety Analysis Report because the increase in allowable stress for occasional operation for short period, of time at higher than design pressure o
conditions makes up for the thinner wall condition. The submittal reported that the i
only Code requirement that cannot be met is the steady state hoop stress limit of i
B31.1, paragraph 104.1.2(a)l. The stress limit is not exceeded at normal operating pressure (850 psig) and pressure at hot standby (900 psig). but the limit is exceeded l
at the piping design pressure of 1000 psig.
Therefore, a linear elastic fracture mechanics analysis was performed, similar to ASME Section XI, Article IWB-3640, 1983 through Winter 1985 Addenda. This analysis was not taken completely from the Code because our pipe material and size are outside the scope of the rules. The results of this analysis demonstrated that there was an adequate margin of safety to operate with the flaw in its present condition. We requested temporary relief at that time, to continue operating.
l Reference (b) reported the completion of a finite element analysis of the piping flaw.
With completion of the finite element analysis we had met the intent of B31.1, We no j
l longer required relief from the Code requirements, based on this statement from the i
Foreword of B31.1-1967, which also exists in the 1989 edition:
The Code never intentionally puts a ceiling limit on conservatism. A designer is free to specify more rigid requirements as he feels they may be justified. Conversely, a designer who is capable of a more rigorous analysis than is specified in the Code may justify a less conservative design and still satisfy the basic intent of the Code.
References (c) and (d) provided additional information on the finite element analysis.
Reference (c) also forwarded an internal memo from our Principal Metallurgist recommending that we not repair the pipe. This recommendation was based on the results of the finite element analysis and a conclusion that any additional welding would result in stresses that would be hard to quantify. This may actually be worse than no repair.
INTENTIONS We have decided to withdraw the Fracture Mechanics Analysis, for the following reasons:
o The finite element analysis fully meets the intent of the original Construction Code for the area of concern and thus clearly demonstrates that relief is no longer necessary. Therefore, we believe that this is the only analysis necessary to close out this issue.
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- Doczm:nt Contr:1 Desk November 15,1989
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o Even the latest edition of ASME Section XI does not include explicit rules for fracture mechanics analysis of the size piping we are dealing with. Therefore, there - is no nationally recognized standard of coefficients to use in the analysis. We do not believe that it is in the best interest of either BG&E or the NRC to continue down unprecedented avenues.
STEAM GENERATOR TUBE RUFTURE PROCEDURE Our current Steam Generator Tube Rupture Procedure, EOP-6, contains direction to prevent over-pressurizing the affected Steam Generator. These directiotis include:
Steaming the affected Steam Generator to the Condenser or Atmosphere, and/or Blowdown to the Condenser or Miscellaneous Waste, and/or Lowering RCS pressure to equalize with affected Steam Generator pressure Specific guidance is provided to perform these actions.
Combustion Engineering (CE) Simulator results for Steam Generator Tube Rupture events with operations in accordance with emergency procedure guidelines do not approach steam generator over-fill or over-pressurization. The Calvert Cliffs simulator and emergency procedure guidelines are very similar to CE's. Therefore, we believe. our procedure prevents steam line over-pressurization for these events.
CONCLUSION We believe this letter addresses the remaining issues related to this matter. - We believe that with this information you will conclude that no further actions need to be taken on our part.
Should you have any further questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours, j
GCC/DLS/bjd cc:
D. A. Brune, Esquire J.
E.
Silberg, Esquire R. A.Capra, NRC S. A.McNeil, NRC W. T. Russell, NRC J.
E. Beall, NRC T. Magette, DNR