ML20217J616

Safety Evaluation Re Alternative to Reactor Pressure Vessel Circumferential Weld Insps for Plant,Unit 3

ML20217J616
Person / Time
Site:	Peach Bottom
Issue date:	10/07/1997
From:	NRC (Affiliation Not Assigned)
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References
NUDOCS 9710210195
Download: ML20217J616 (4)

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NUOLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001

%...../ l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE TO REACTOR PRESSURE VESSEL (RPV) CIRCUMFERENTIAL WELD INSPEC PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION. UNIT 3 DOCKET NO. 50-278

1.0 INTRODUCTION

By letter dated September 4, 1997, as supplemented by letter dated September 22, 1997, PECO Energy Company (PECO, the licensee) requested NRC approval of Unit 3 RPV alternative reactor vessel weld examinations pursuant to the provisions of 10 CFR 50.55a(g)(6)(ii)(A)(5) for the next two operating cycles. The licensee proposed these inspections as an alternative to the augmented examinations specified in 10 CFR 50.55a(g)(6)(li)(A)(2) for circumferential welds, and as an alternative to the inservice inspection requirements for circumferential welds specified in the American Society of Mechanica' Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1980 Edition, with Winter 1981 Addenda. The requested alternative inspections are consistent 97-63, " Status ofwith hRCinformation Staff Reviewcontained of BWRVIP-05."in NRC Information Notice (IN)

Section 50.55a(g)(G)(ii)(A) to Title 10 of the Code of federal Regulations

((10 CFR 50.55a(g)(6)(ii shell weld examination as)(A)) requires that licenseet perform an expanded RPV

" expedited" basis. specified in the 1989 Edition of Section XI, on an

" Expedited," in this context, effectively meant during the inspection interval when the Rule was approved, or the first period of the next inspection interval. The final Rule was publisned in the Federal Register on August 6, 1992 (57 FR 34666). The NRC staff incorporated the 1989 Edition of the ASME Code into the regulations, requiring that licensees perform volumetric examination of " essentially 100 percent" of the RPV pressure-retaining shell welds during all inspection intervals. Section 50.55a(a)(3)(i) ((10 CFR 50.55a(a)(3 1 indicates that alternatives to the requirements in 10 CFR 50.55a(g) are)j(us))ified t when the proposed alternative provides an acceptable level of quality and safety.

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), a technical committee of the BWR Owners Group (BWROG), submitted the proprietary report, "BWR Vessel and Internals Project, BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," in a September 28, 1995, letter, as supplemented by June 24, and October 29, 1996, letters, and May 16, June 4, and June 13, 1997, letters. The proprietary report proposed to reduce the BWR RPV weld 9710210193 971007 PDR ADOCK 05000278 G PDR

2-scope of inspection from essentially 100 percent of all RPV shell wel'ds to 50 percent of the axial welds and 0 percent of the circumferential welds. The BWRVIP modified their proposal in an t.ctober 29, 1996, letter to increase axial weld examination to 100 percent from 50 percent, while still proposing to inspect essentially 0 percent of the circumferential RPV shell welds.

However, the intersection of the axial and circumferential welds would have included approximately 2-3 percent of the circumferential welds.

On May 12, 1997, the NRC staff and members of the BWRVIP met with the Commission to discuss the NRC staff's review of the BWRVIP-05 report. The NRC staff has initiated a broader, risk-informed review of the BWRVIP-05 proposal in accordance with Commission guidance provided in May 30, 1997, Staff Requirements Memorandum (SRM) M9705128.

In IN 97-63, the NRC staff indicated that it would consider technically-justified alternatives to the augmented examination in accordance with 10 CFR 50.55a(a)(3)(i), 10 CFR 50.55a(a)(3)(ii), and 30.55a(g)(6)(ii)(A)(5), from BWR licensees who are scheduled to perform inspections of the BWR RPV circumferential welds during the fall 1997 or spring 1998 outage seasons. The NRC staff would ccnsider acceptably-justified alternatives for inspection delays of up to 40 months or two operating cycles (which ever is longer) for BWR RPV circumferential shell welds only.

2.0 BACKGROUND

- NRC STAFF BWRVIP-05 REPORT ASSESSMENT The NRC staff documented its independent assessment of the BWRVIP-05 proposal in an August 14, 1997, letter to Carl Terry, BWRVIP Chairman. The NRC staff concluded that the industry's assessment does not sufficiently address risk, and additional work is necessary to provide a complete risk-informed evaluation.

The NRC staff performed its assessment for BWR RPVs fabricated by Chicago Bridge and Iron (CB&I), Combustion Engineering (CE), and Babcock & Wilcox (B&W). The NRC staff assessment identified cold over-pressure events as ceing the limiting transients that could lead to BWR RPV failure. The NRC staff determined the conditional probability of failure for axial and circumferential welds fabricated by CB&I, CE, and B&W, using the pressure and temperature resuliing from a cold over-pressure event in a foreign reactor, and the parameters identified in Table 7-1 of the NRC staff's independent assessment. Table 7-9 of the NRC staff's assessment identifies the conditional ~ probability of failure for the reference cases, and the 95 percent confidence uncertainty bound cases for axial and circumferential welds fabricated by CB&I, CE and B&W. B&W fabricated vessels wera determined to nave the highest conditional probability of failure. The input material parameters used in the reference case analysis for B&W fabricataa vessels resulted in a reference temperature (RT ,) at the vessel inner surface of Il4.3*F. In the uncertainty analysis, b neutron fluence evaluation had the greatest RT value (145'F) at the inner surface. Vessels with RT values lessthanth37oseresultingfromtheNRCstaff'sassessmentwillhave. bess

4 embrittlement than the vessels simulated in the staff's assessment, and should have a conditional probability of vessel failure less than or equal to the values in the staff's assessment.

Weld failure probability is the product of the critical event frequency and the conditional probability of the weld failure for that event. The best-estimate failure frequency from the NRC staff's assessment is 6.0 X 10

• per reactor year, and the uncertainty bound failure frequency is 3.9 X 10 s per reactor year, using the event frequency for a cold over-pressure w ent and the conditional probabilit,' of vessel failure for B&W fabricated circumferential welds.

3.0 LICENSEE'S TECHNICAL JUSTIFICATION PECO's September 22, 1957, letter indicated that they will be performing an examination of the reactor vessel longitudinal shell welds to the maximum extent practical from the inner diameter, within the constraints of vessel internal . restrictions. The extent of weld examination coverage anticipated l for the longitudinal shell welds is identified on Table 1 of G. A. Hunger's, i April 29, 1997, letter to the NRC. The licensee indicated that their current-l examination plan was designed to provide longitudinal weld coverage, however.

l incidental cuverage will result in an estimated portion of 2-3 percent of the intersecting horizontal weld.

The licensee also indicated in the September 4,1997, letter that the basis for requesting alternative inspections is the BWRVIP-05 report. The report indicated that the probability of failure of BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. The NRC staff's independent assessment of the BWRVIP-05 report also concluded this. TheBWRVIP-05reportindicatestyt,foratypicalBWRRPV,thefailure probability for axial welds is 2.7 X 10' , ar.d the failure probability for circumferential welds is 2.2 X 10' for 40 years o' plant operation.

The licensee calculated the RT value for the limiting PBAPS 3 circumferential weld at the en77of the requested alternative inspections period using the methodology in Regulatory Guide (RG) 1.99, Revision 2. The RT m values calculated in accordance with RG 1.99, Revision 2, depend upon the neutron fluence, the amounts of copper and r.ickel in the circumferential weld, and its :nirradiated RT m. The licensee determined the maximum neutron fluence at the end of the requested alternative inspections period at the innep surface of the limiting circumferential beltline weld to be 0.045 X 10" n/cm. The amounts of copper and nickel in the limiting circumferential beltline weld are 0.11 percent and 0.96 percent, respectively. The plant-specific unirradiated RT m for the limiting circumferential beltline weld is -50*F. Using these parameters and the methodology in Regulatory Guide 1.99, Revision 2, the licensee determined that the RT value for the circumferential weld at the end of the alternative in,spections period is 30.8'F. This is less than the reference case for the B&W fabricated vessels

in the NRC' staff's assessment. The licensee utermined that the conc'lusions of the BWRVIP-05 report are bounded for the PBAPS 3 RPV, since the R7 of the PBAPS staff's 3 beltline circumferential weld is less than the values in Ee NRC assessment.

The licensee assessed the systems that could lead to a cold over-pressurization of the PBAPS 3 RPV. These included the high pressure coolant injection, reactor core isolation cooling, r.ormal feedwater, control rod drive and reactor water cleanup systems. In all cases, the operators are trained-in methods of controlling water level within specified limits ir addition to responding to abnormal- water level conditions during shutdown. The licensee har established plant-specific procedures to provide guidance to the operators r

regarding compliance with the Technical Specification pressure-temperature limits. The licensee determined that a non-design basis cold over-pressure transient is unlikely to occur during the next two operating periods, on the basis of ti.9 pressure lindts of the operating systems, operator training, and l

established plant-specific procedures. Therefore, the licensee-concluded that the probabi'ity of a cold over-pressure transient is considered to be less than or equal to that used in the NRC staff's assessment.

4.0 NRC STAFF REVIEW OF LICENSEE'S TECHNICAL JUSTIFICATION The NRC staff r.cnfirmed that the RT , value for the circumferential weld at the'end of the alternative inspections perind is less than the values in the reference case and uncertainty analysis for the B&W fabricated vessels.' RT is a measure of the amount of irradiation embrittlement. ThePBAPS3RPVwU1 have less embrittlement than the B&W fabricated. vessels, and will have a

-conditional probability of-vessel failure less.than or equal to that estimated in the NRC staff's assessment. This is due to the RT the value in the reference case and the values in the.,uncertainty value being less than analysis _for B&W fabricated vessels. The probability of a cold over-pressure transient should be minimized during the next two. operating

-limits on the operating systems, and the licensee' periods, based s operator on pressure training and established procedures.

5.0 CONCLUSION

S

1) The conditional probability of vessel- failure should Le less than or equal to that e.stimated from the NRC staff's assessment, based ~on the licensee's assessment of the materials in the circumferential weld in the beltline of:

the PBAPS 3'RPV.

2) -The probability of cold over-pressure transients should be minimized during the.next two operating periods, based on the licensee's operator training and established procedures.

3) The NRC staff concludes that the PBAPS 3 RPV can be operated during.the next two operating periods with an acceptable level of quality and safety, and the inspection of the circumferential welds can be delayed for two operating periods, based on the previous two conclusions.

Principal Contributor: K. Karwoski Date: October 7, 1997