ML20205Q529

SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained

ML20205Q529
Person / Time
Site:	Quad Cities
Issue date:	04/16/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20205Q521	List: ML20205Q519 ML20205Q529
References
GL-88-01, GL-88-1, NUDOCS 9904210292
Download: ML20205Q529 (2)

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UNITED STATES

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WASHINGTON, D.C. 20686-0001 ps...../

l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO FLAW EVALUATION OF RECIRCULATION LINE WELD 02BS-F4 COMMONWEALTH EDISON COMPANY 6_N_Q MIDAMERICAN ENERGY COMPANY OUAD CITIES NUCLEAR POWER STATION. UNIT 1 DOCKET NO. 50-254

1.0 INTRODUCTION

During the recent refueling outage for Unit 1 (Q1R15), Commonwealth Edison Company (Comed, the licensee) performed inservice inspection on the recirculation system piping using ultrasonic examination in accordance with Generic Letter (GL) 88-01, *NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping." In a letter dated November 24,1998, the licensee reported that seven welds were found with flaw indications that exceeded the acceptance -

criteria in subarticle IWB-3500, " Acceptance Standards," of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),1989 Edition.

.Additionalinformation was provided by the licensee in conference calls held on November 19, 24 and 30,1998. Welds 02AD-F12 and 02AS-S4, exhibiting indications identified in the preceding outage, showed no change in flaw size and the previous flaw evaluation remains applicable and acceptable for continued operation for the next operating cycle. One new flaw indication was detected on weld 02BS-F4. A flaw evaluation was performed on this weld to justify continued operation without repair. Each of the four remaining indications on welds 02AD-F8,02BS-F7,02AS-F9, and O2BS-F14 had a weld overlay repair applied in accordance with ASME Code Case N-504," Alternative Rules for Repair of Class 1,2, and 3 Austenitic Stainless Steel Piping,Section XI, Division 1."

The results of the licensee's flaw evaluation have shown that Quad Cities Nuclear Power Station, Unit 1, can be safely operated for the next operating cycle of 24 months because the subject indication meets the criteria of IWB-3640,Section XI of the ASME Code.

2.0 EVALUATION l

l The licensee performed a flaw eva!uation on weld 02BS-F4 based on the indication detected by

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ultrasonic examination. The fracture mechanics evaluation was conducted using the procedures provided in Appendix C and Paragraph IWB-3640 of the ASME Code,Section XI.

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' The crack growth evaluation to determine the projected crack depth was conducted using the procedures outlined in NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," Revision 2.

Weld 02BS F4 is an induction heating stress improvement (IHSI) treated flux weld fabricated by the shielded metal-arc welding process. The flaw was located on the pump suction side of the l

recirculation piping. The pipe nominal size is 28 inches and the material is Type 304 stainless steel. The fuel cycle length was assumed to be 24 months or approximately 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br />. The indication depth, indication length and pipe thickness for weld 02BS-F4 are 0.25, 27.0, and 1.24 inches, respectively. The limit load analysis method, based upon the procedures outlined in Section XI of the ASME Code, was used to calculate the primary membrane and bending stresses. The safety factor value is 2.77 for normal / upset conditions and 1.3g for emergency / faulted conditions. To account for the lower fracture toughness attributed to flux welds a Z factor is used which is a function of the nominal pipe size. The applied piping stresses were calculated from the reported axial and bending loads at the subject weld.

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The crack growth due to fatigue is negligible. Therefore, only crack growth from intergranular stress corrosion cracking is considered in the flaw evaluation. Weld 02BS-F4 was IHS! treated in order to eliminate the tensile weld residual stress pattern and produce a compressive residual stress pattern at the inside diameter surfaces of the girth welds. To assure a conservative approach, this beneficial effect of IHSI was not considered in this evaluation and as-welded residual stress distribution was used. The as-welded residual stress distribution is determined from a polynomial equation provided in NUREG-0313, Revision 2. The applied stress was calculated by summing the pressure stress and the membrane and bending stresses from the weight and thermal loadings. The internal pressure was assumed to be 1000 psi. Stresses from the most limiting condition were considered. The licensee's crack growth calculations are based on an equation provided in NUREG-0313, Revision 2, for plants with normal water chemistry. Therefore, the licensee's crack growth evaluation is quite conservative considering the plant is going to be operated with hydrogen water chemistry. The limit load evaluation for weld O2BS-F4 allows a maximum crack depth of 0.793 inch (0.6t, where t is the wall thickness) for a 360 degrees circumferential flaw. The predicted flaw size at the end of the next operating cycle (24 months) has a depth of 0.472 inch and length of 50.971 inches which are well within j

the Code allowable values. The results of the licensee's crack growth evaluation are acceptable because it followed the guidelines established in NUREG-0313, Revision 2, and met I

the Code acceptance criteria.

3.0. CONCLUSION Based on a review of the licensee's submittal, the staff concludes that Quad Cities Nuclear Power Station, Unit 1, can be safely operated for the next fuel cycle with weld O2BS-F4 in its current condition because the structural integrity of the weld will be maintained. However, I

continued plant operation beyond the next fuel cycle will depend on the satisfactory evaluation of the re-inspection results or implemaiting acceptable repairs during the next refueling outage.

Principal Contributor: W. Koo Date: April 16,1999 t

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