ML20196A506
| ML20196A506 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/23/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20196A502 | List: |
| References | |
| NUDOCS 9811300019 | |
| Download: ML20196A506 (2) | |
Text
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lp;#""'%f.1 UNITED STATES NUCLEAR REGULATORY COMMISSION U
WASHINGTON, D.C. 2000641001 i
SAFETY EVALUATION BY TI-lE OFFICE OF NUCI PAR REACTOR REGULATION p
FLAW EVALUATION OF MAIN STEAM NO771 F TO SHFlI WFI r)
COOPER NUCLEAR STATION '
i NEBRASKA PUBLIC POWER DISTRICT i
DOCKET NO. 50-298 l'
L
1.0 INTRODUCTION
r During the fall 1998 refueling outage at Cooper Nuclear Station, a flaw indication was found by ultrasonic examination in the main steam nonle to shell weld NVE-BD-N3A.
- The main steam nonles are 24 inches in diameter and welded to the vessel made of 6-inches thick cylindrical shell. The indication was reported to be a subsurface planar.
indication with a through wall dimension of 0.88 inch, a length of 12.25 inches and a surface separation of 2.56 inches. The flaw indication is believed to be a fabrication defect, probably due to a slag inclusion formed during welding. The size of the indication exceeds the acceptance criteria of ASME Code Section XI, IWB-3512-1 and requires fracture mechanics evaluation in accordance with the procedures and acceptance criteria in IWB-3600 of Code Section XI. Nebraska Public Power District's (the licensee) fracture mechanics evaluation was provided in its letter dated November 17,1998.
2.0 EVALUATION j
J The licensee performed a fracture mechanics evaluation and a primary stress evaluation I
for the following operating conditions: Hydro test, Normal (Level A), Upset (Level B),
Emergency (Level C), and Faulted (Level D). The stresses at the location of the indication i
are primarily resulting from intemal pressure and thermal gradient. In the fracture mechanics evaluation, the licensee assumed the RT,er of the subject weld to be 18' F, because the fracture toughness data for the subject nonle weld could not be located.
The licensee stated that the highest RT,or f the steam outlet nonle forgings was 18'F, o
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, the highest RT,er of the upper shell plate was 14*F, and the other welds for which the I
data is available have RT,er values in the neighorhood of-50*F. The staff considers that
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the licensee's assumption is acceptable because the assumed RT,e7 or the subject f
nonle weld is significantly higher than that of the other welds. However, the licensee should continue the effort to locate the actual test data. In the fatigue crack growth calculations,120 cycles of start-up/ shut down (design life) were assumed. The assumed fatigue cycles are conservative because the subject evaluation is performed to support the operation of one fuel cycle, not the life of the plant. The results of the licensee's fracture 9811300019 981123 PDR ADOCK 05000298 ENCLOSURE P
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mechanics evaluation showed that the applied stress intensity factor value at the location of the subject indication is lower than the allowable stress intensity factor value with required Code safety. margin for each of the evaluated operating conditions. The results of the licensee's maximum primary membrane stress evaluation showed that the primary stresses at the location of the indication will not exceed 1.5 S, at the end of next fuel
. cycle. This meets the primary stress requirements of Section lli of the ASME Code.
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- The staff has determined that the results of the licensee's fracture mechanics and pnmary j
stress evaluation are acceptable because the methodologies used in the licensee's flaw l
evaluation follow the Code procedures and its results meet the respective Code i
requirements. Thcrefore, it is not necessary to repair the subject flaw for continued operation of one fuel cycle.
3.0 CONCLUSION
Based on the staff's review of the licensee's fracture mechanics evaluation, the staff concludes that the Cooper Nuclear Station can be safely operated for at least one fuel cycle with the indication in as-is condition, since the licensee's fracture mechanics evaluation has provided reasonable assurance that the structural integrity of the subject nozzle weld can be maintained. However, for continued operation beyond the next fuel cycle, it should be supported by the satisfactory results of reinspection or repair of the subject nozzle weld in accordance with the Code requirements.
Principal Contributor: W. Koo Date: November 23,.1998
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