ML20198J092

Safety Evaluation Authorizing Licensee & Suppls & 16 Request for Approval of Alternative Reactor Vessel Weld Exam,Per 10CFR50.55a(g)(6)(ii)(A)(5) for Plant, Unit 2 for Next 2 Operating Cycles

ML20198J092
Person / Time
Site:	Brunswick
Issue date:	09/18/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198J088	List: ML20198J086 ML20198J092
References
FACA, NUDOCS 9709290195
Download: ML20198J092 (5)

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g } NUCLEAR RE2ULATORY COMMISSION g $ C'ASHINGTON D.C. 30666 4 001

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGUL ATION RELIEF FROM CERTAIN REACTOR PRESSURE VESSEL INSPECTION REQUIREMENTS I

BRUNSWICK STEAM ELECTRIC PLANT UNIT N0. 2 CAROLINA POWER & LIGHT COMPANY I

DOCKET N0: 50-324

1.0 INTRODUCTION

By letter dated August 22. 1997, as supplemented by letters dated Seatember 2. 1997, and September 16, 1997. Carolina Power & Light Company (C)&L or the licensee) requested an alternative to performing the reactor 3ressure vessel (RPV) circumferential shell weld examination requirements of

)oth the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).Section XI, 1980 Edition, with Winter 1981 Addenda (inservice inspection). and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) for the Brunswick Steam Electric Plant. Unit No. 2 (BSEP2) for two operating cycles. The alternative was proposed pursuant to the provisions of 10 CFR 50.55a(g)(6)(ii)(A)(5) and is granted under the provisions of 10 CFR 50.55a(a)(3)(1). The alternative is consistent with information contained in Infonnation Notice (IN) 97-63. " Status of NRC Staff Review of BWRVIP-05." The Seatember 2. 1997, letter contained supplemental information requested by the NRC staff during a telephone conference call with the licensee on August 26, 1997, related to plant procedures and operator training. The September 16, 1997. letter provided clarification regarding the regulacory basis for the request and the proposed alternative.

The alternativ proposed by CP&L is the performance of inspections of essentially 100 percent of the BSEP 2 RPV shell longitudinal seam welds and essentially 0 percent of the RPV shell circumferential seam welds during Refueling Outage 12 (B213R1). which will result in partial examination (2 - 3 percent) of the circumferential welds at or near the intersections of the longitudinal and circumferential walds.

The requirement for inservice inspections, which include RPV circumferential weld inspection. derives from the Technical Specifications (TS) for BSEP2 which state that the inservice inspection (ISI) and testing of the ASME Code Class 1, 2. and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g). Pursuant to 10 CFR 50.55a(g)(4). ASME Code Class 1, 2.

and 3 components shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code.Section XI. " Rules for Inservice Inspection of Nuclear Power Plant Components." to the extent practical within the limitations of design.

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geometry, and materials of construction of Lhe components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsecuent intervals comply '

with the requirements in the latest edition and addenca of the ASME Code.

Section XI incor> orated by reference in 10 CFR 50.55a(b) on the date 12 months prior to t1e start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable ASME Code.

Section XI. for Brunswick Units 1 and 2. during the second 10-year ISI interval is the 1980 Edition through the Winter 1981 Addenda.

Section 50.55a(g)(6)(ii)(A) to Title 10 of the Code of Federa? Regulations (10 CFR 50.55a(g)(6)(ii)(A)) requires that licensees perform an expanded RPV shell weld examination as specified in the 1989 Edition of Section XI of the ASME Code on an " expedited" basis. " Expedited." in this context, effectively meant during the inspection interval when the Rule was approved or the first i period of the next inspection interval. The final Rule was published in the  !

i federa? Register on August 6, 1992 (57 FR 34666). By incorporating into the i

regulations the 1989 Edition of the ASME Code, the NRC staff required 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)(1) (10 CFR 50.55a(a)(3)(1)) indicates that alternatives to the requicements in 10 CFR 50.55a(g) are justified when the proposed alternative provides an acceptable level of quality and safety.

By letter iated September 28, 1995, as supplemented by letters dated June 24 and October 29, 1996, and May 16. June 4. and June 13. 1997, 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)." which proposed to reduce the scope of inspection of the BWR RPV welds from essentially 100 percent of all RPV shell welds to 50 percent of the axial welds and 0 percent of the circumferential welds. By letter dated October 29, 1996, the BWRVIP modified their proposal to inc~r ease the examination of the axial welds to 100 percent from 50 percent while still proposing to inspect essentially 0 percent of the circumferential RPV shell welds, except that 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. In accordance with guidance provided by the Commission in Staff Requirements Memorandum (SRM) M9705128. dated May 30, 1997, the staff has initiated a broader. risk-informed review of the BWRVIP-05 proposal.

In IN 97-63. the staff indicated that it would consider technically -justified alternatives to the augmented examination in accordance with 10 CFR 50.55a(a)(3)(1) 10 CFR 50.55a(a)(3)(ii). and 50.55a(g)(6)(11)(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.

Acceptably justified alternatives would be considered for inspection delays of up to 40 months or two operating cycles (whichever is longer) for BWR RPV circumferential shell welds only.

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2.0 BACKGROUND

- STAFF ASSESSMENT OF BWRVIP-05 REPORT:

The staff's independent assessment of the BWRVIP-05 proposal is documented in a letter dated August 14, 1997, to Mr. Carl Terry. BWRVIP Chairman. The 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 staff's assessment was performed for BWR RPVs fabricated by Chicago Bridge and Iron (CB&I). Combustion Engineering (CE), and Babcock & Wilcox (B&W). The staff assessment identified cold over-pressure events as the limiting transients that could lead to failure of BWR RPVs. Using the pressure and temperature resulting from a cold over-pressure event in a foreign reactor and the parameters identified in Table 7-1 of the staff's independent assessment, the staff determined the conditional 3robability of failure for axial and circumferential welds fabricated by C3&l CE, and B&W. Table 7-9 of the staff's assessment identifies the conditional probability of failure for the reference cases and the 95 percent confidence uncertainty bound cases for axial end circumferential welds fabricated by CB&I. CE and B&W. B&W fabricated vessels were determined to have the highest condidonal probability of failure. The input material parameters used in the analysis of the reference case for B&W fabricated vessels resulted in a reference temperature (RTer) at the vessel inner surface of 114.5*F. In the uncertainty anal" sis, the neutron fluence evaluation had the greatest RTer value(145F)attbe inner surface. Vessels with RTer values less than those resulting from the staff's assessment will have less 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. 1' The failure probability for a weld is the product of the critical event frequency and the conditional probability of the weld failure for that event.

Using the event frequency for a cold over-pressure event and the conditional 3robability of vessel failure for B&W fabricated circumferential welds, jhe-3est-estimate failure frequency from the staff's assessment is 6.0 X 10' per reactor year and the uncertainty bound failure frequency is 3.9 X 10 4 per '

reactor year.

3.0 LICENSEE TECHNICAL JUSTIFICATION:

The licensee indicated in the August 22, 1997, letter that the basis for requesting the alternative inspections is the BWRVIP-05 report, which stated that the probability of failure of BWR RPV circumferential shell welds is <

orders of magnitude lower than that of the axial shell welds. This conclusion was also demonstrated in the staff's inde)endent assessment of the BWRVIP-05 report. The BWRVIP-05 report indicates t1at, for a typical BWR RPV the failure probability for axial welds is 2.7 X 10' and the failure probability for circumferential welds is 2.2 X 10* for 40 years of plant operation.

The licensee calculated the RT the end of the requested reliei1period value using for thethe BSEP2 circumferential methodology weld at in Regulatory Guide (RG) 1.99. Revision 2. The rte 1 values calculated in accordance with RG

4 1.99 Revision 2. depend upon the neutron fluence, the amounts of copper and nickel in the circumferential weld, and its unirradiated RTc. The licensee determined the maximum neutron fluence at the end of the next two operating cygles at the inner surface of the circumferential beltline weld to be 0.061 X 10 . The amounts of copper and nickel in the circumferential beltline weld are 0.02 percent and 0.90 percent, respectively. The plant-specific unirradiated RT for the circumferential beltline weld is 10 F. Using these parameters and ,the methodology in Regulatory Guide 1.99. Revision 2. the licensee determined that the RT endofthereliefperiodis27.E*valueforthecircumferentialweldatthe F. which is less than the reference case for the B&W fabricated vessels in the staff's assessment. Since the RT of the BSEP2 beltline circumferential weld is less than the values in the , staff's assessment, the licensee concluded that the conclusions of the BWRVIP-05 report are bounded for the BSEP2 RPV.

The licensee assessed the systems that could lead to a cold over-pressurization of the BSEP2 RPV. These included the high pressure coolant injection, reactor core isolation cooling, standby liquid control, control rod drive and reactor water cleanup systems. In all cases, the operators are trained in methods of controlling water level within specified limits in addition to responding to abnormal water level conditions during shutdown.

Plant-specific procedures have been established to provide guidance to the operators regarding compliance with the Technical Specification pressure-temperature limits. On the basis of the pressure limits of the operating systems, operator training, and established plant-specific procedures the licensee determined that a non-design basis cold over-pressure transient is unlikely to occur during the next two opercting cycles. Therefore, the licensee concluded that the probability of a cold over-pressure transient is considered to be less than or equal to that used in the staff's assessment.

4.0 STAFF REVIEW 0F LICENSEE TECHNICAL JUSTIFICATION:

The staff confirmed that the RT, value for the circumferential weld at the end of the relief period 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. Since the RT valu,e is less than the value in the reference case and the values in the , uncertainty analysis for B&W fabricated vessels, the BSEP2 RPV will 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 staff's assessment.

Based on pressure limits on the operating systems, and the licensee's operator training and established procedures, the probability of a cold over-pressure transient should be minimized during the next two operating periods.

5.0 CONCLUSION

S:

Based upon its review, the staff reached the following conclusions:

1) Based on the licensee's assessment of the materials in the circumferential weld in the beltline of the BSEP2 RPV. the conditional

5-probability.of vessel failure should be less than or equal to that-estimated from the staff's assessment.

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

3) Based on the previous two conclusions, the staff concludes that BSEP2 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.

Therefore, the proposed alternative to performing the RPV examination recuirements of the ASME Code,Section XI,1980-Edition, with Winter 1981 Adcenda, and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) at BSEP2 for circumferential shell welds for two operating cycles is authorized pursuant to 10 CFR 50.55a(a)(3)(1).

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