ML20217E747

Safety Evaluation Accepting Alternative to Insp of Reactor Pressure Vessel Circumferential Welds

ML20217E747
Person / Time
Site:	Brunswick
Issue date:	04/21/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217E713	List: ML20217E710 ML20217E747
References
NUDOCS 9804270396
Download: ML20217E747 (6)

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[ psp"c% 1 UNITED STATES g .) ,j NUCLEAR REGULATORY COMMISSION o o t WASHINGTON, D.C. 20565-0001

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

i ALTERNATIVE TO INSPECTION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELDS j BRUNSWICK STEAM ELECTRIC PLANT UNIT N0. 1 CAROLINA POWER & LIGHT COMPANY l

DOCKET N0: 50-325

1.0 INTRODUCTION

l By [[letter::BSEP-97-0459, Requests Approval to Use Alternative Requirements for ISI Delineated in ASME Section XI Code Case N-535.Alternative Is Needed to Extend Third Period of Second ten-yr ISI Interval to Coincide W/End of Bsep,Unit 1,refueling Outage 11|letter dated November 17, 1997]]. Carolina Power & Light Company (CP&L or the licensee) requested an alternative to performing the reactor pressure vessel

) (RPV) circumferential shell weld examinations requirements of both the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME l

Code).Section XI, 1980 Edition, through the 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 (BSEP)

Unit 1. The alternative was proposed pursuant to the provisions of 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(5), and is consistent with information contained in Information Notice (IN) 97-63 " Status of NRC Staff Review of BWRVIP-05." l The alternative proposed by CP&L is the performance of inspections of essentially 100 percent of the BSEP Unit 1 RPV shell longitudinal seam welds and essentially 0 percent of the RPV shell circumferential seam welds during Refueling Outage 11, which will result in partial examination (2 - 3 percent) of the circumferential welds at or near the intersections of the longitudinal and circumferential welds.

The requirement for inservice inspections, which include RPV circumferential weld inspection, derives from the Technical Specifications (TS) 4.0.5 for BSEP Unit 1 which state that "the inservice inspection (ISI) and testing of the

, ASME Code Class 1, 2 and 3 components be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, and aoplicable addenda, as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC. .~. Pursuant to the requirements of 10 CFR 50.55a(g)(4).

ASME Code Class 1.2. and 3 components shall meet the requirements, except the 9804270396 900421 PDR ADOCK 05000325 P PDR

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• 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. geometry, and materials of construction of the components. The l regulations require that inservice examination of components and system l pressure tests conducted during the first 10-year interval and subsequent l intervals comply with the requirements in the latest edition and addenda of l the ASME Code.Section XI. incorporated by reference in 10 CFR 50.55a(b) on l the date 12 months prior to the start of the 120-month interval, subject to l the limitations and modifications listed therein. The applicable ASME Code'.

Section XI. for Brunswick. Unit 1. during the current 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 Federal hegulations l

(10 CFR 50.55a(g)(6)(ii)(A)) requires that licensees perform an expanded RPV l shell weld examination as specified in the 1989 Edition of Section XI of the l ASME Code. on an " expedited" basis. " Expedited." in this context. effectively l meant during the inspection interval when the Rule was approved or the first

! period of the next inspection interval. The final ~ Rule was published in the Federal Register on August 6, 1992 (57 FR 34666). By incorporating into the 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)(i) (10 CFR 50.55a(a)(3)(i)) indicates that alternatives to the requirements in 10 CFR 50.55a(g) are justified when the proposed alternative provides an acceptable level of quality and safety.

By letter dated 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 increase the 1 examination of the axial welds to 100 percent from 50 percent while still I 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) M970512B, 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)(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.

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.

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 overpressure events as the limiting transients that could lead to failure of BWR RPVs. Using the pressure and temperature resulting from a cold overpressure event in a foreign reactor and the parameters identified in Table 7-1 of the staff's independent assessment, the staff determined the conditional probability of failure for axial and 1 circumferential welds fabricated by CB&I. CE, and B&W. Table 7-9 of the I 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 were determined to have the highest conditional 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 analysis, the neutron fluence evaluation had the greatest RTer value (145 F) at the inner surface. Vessels with rte 1 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, l

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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 overpressure event and the conditional probability of vessel failure for CB&I fabricated circumferential welds, the best-estimate failure frequency from the staff's assessment is 6.0 X 10-inn per reactor year and the uncertainty bound failure frequency is < 2.8 X 10'*D per reactor year.

3.0 LICENSEE TECHNICAL JUSTIFICATION:

! The licensee indicated in the [[letter::BSEP-97-0459, Requests Approval to Use Alternative Requirements for ISI Delineated in ASME Section XI Code Case N-535.Alternative Is Needed to Extend Third Period of Second ten-yr ISI Interval to Coincide W/End of Bsep,Unit 1,refueling Outage 11|November 17, 1997, letter]] that the basis for requesting the alternative inspections is the BWRVIP-05 report, which stated that the prcbability 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 independent assessment of the BWRVIP-05 report. The BWRVIP-05 report indicates that, for a typical BWR RPV the

, failure probability for axial welds is 2.7 X 10 and the failure probability 7

for circumferential welds is 2.2 X 1042 for 40 years of plant operation.

The licensee calculated the RTm value for the BSEP Unit 1 circumferential weld at the end of the requested relief period using the methodology in Regulatory Guide (RG) 1.99. Revision 2. The RT, values calculated in accordance with RG 1.99 Revision 2. depend upon the neutron fluence, the amounts of copper and nickel in the circumferential weid, and its unirradiated RTuor. The licensee determined the highest neutron fluence at the end of the next two operating cveles at the inner surface of the circumferential beltline weld to be 0.063 X 10' n/cm2 . The amounts of copper and nickel in the limting circumferential beltline weld is 0.06 percent and 0.87 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, value for the circumferential weld at the end of the relief period is 64.26 F. The licensee noted that the RTm resulted from the plant-specific unirradiated RTm value that was provided in CP&L's letter dated November 16. 1995, and that use of the generic initial RT, value of -56 *F would yield an RT, value of 14.63 F. The larger RT, value that results from plant-specific unirradiated RT, is still less than the most limiting reference case (B&W fabricated vessels) in the staff's assessment. Since the RT, of the BSEP. Unit 1. beltline circumferential weld is.less than the limiting RTm value in the staff's Insufficient or no failures to accurately determine reference case failure probability.

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l 5-independent assessment, the licensee concluded that the BSEP, Unit 1, vessel circumferential welds are bounded by the staff's independent' assessment, thus providing additional assurance that the vessel welds are bounded by the BWRVIP-05 report.

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! The licensee assessed the systems that could lead to a cold overpressurization l of the BSEP, Unit 1. RPV. These included the high pressure coolant injection i

(HPCI), reactor core isolation cooling (RCIC), standby liquid control (SLCS),

control rod drive (CRD) and reactor water cleanup systems (RWCU). Both the HPCI and RCIC pumps are steam driven and do not function during cold shutdown.

l The licensee stated that there were no automatic starts associated with SLCS.

Operator initiation of SLCS should not occur during shutdown; however, the j SLCS injection rate is approximately 41 gpm which woul.d allow the operators i sufficient time to control reactor pressure if manual initiation occurred.

The CRD and RWCU systems t.se a feed and bleed process to control RPV level and pressure during normal cold shutdown conditions. The CRD pumps injection rate is less than 60 gpm which allows sufficient time for operators to react to l unanticipated level changes.

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. The licensee also stated that procedural controls for reactor temperature, level, and pressure are an integral part of operator

! training. 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 overpressure transient is unlikely to occur during the requested delay. Therefore, the l

licensee concluded that the probability of a cold overpressure transient is '

considered to be less than or equal to that used in the staff's assessment of BWRIVP-05.

4,0 STAFF REVIEW 0F LICENSEE TECHNICAL JUSTIFICATION:

BSEP Unit 1 is a CB&I fabricated vessel, and the staff noted that the RTuor i value determined from the plant-specific unirradiated RTuo7 (64.26 F) is approximately 14 F higher than the limiting value determined in the staff's assessment for CB&I fabricated vessels (50 *F). However, RTuor is a measure of the amount of irradiation embrittlement, and since CB&I fabricated vessels have very low copper values, they have low amounts of irradiation embrittlement. For comparison, the staff confirmed that the RTuo7 value for

I the circumferential welds at the end of the relief period is less than the values in the limiting reference case and uncertainty analysis for the B&W fabricated vessels. Since the RT,m values are well below the values in the reference case and the uncertainty analysis for B&W fabricated vessels, the l

BSEP Unit 1 RPV will have less embrittlement than the B&W fabricated vessels and will have conditional probability of vessel failure less than or equal to that estimated in the staff's assessment.

The staff reviewed the information provided by the licensee regarding the BSEP. Unit 1. high pressure injection systems, operator training, and plant-specific procedures to prevent RPV cold overpressurization. The information provided sufficient basis to support approval of the alternative examination l request. Based on the high pressure injection systems analyses, operator

! training, and plant-specific procedures, the probability of a cold l

overpressurization transient occurring at BSEP. Unit 1. during the requested delay is low, which is consistent with the staff's assessment.

5.0 CONCLUSION

S:

i l Based upon its review. the staff reached the following conclusions: i

1) Based on the licensee's assessment of the materials in the circumferential weld in the beltline of the Brunswick. Unit 1. RPV. the l conditional probability of vessel failure should be less than or equal to that estimated from the staff's assessment of the limiting case of l B&W fabricated vessels.

2) Based on the licensee's high pressure injection systems analyses.

operator training, and plant-specific procedures, the probability of I

cold overpressure transients should be sufficiently low during the requested delay period.

3) Based on the previous two conclusions, the staff concludes that Brunswick. Unit 1. RPV can be operated during the requested delay period 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 requirements of the ASME Code.Section XI 1980 Edition, with Winter 1981 Addenda, and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) at Brunswick Unit 1 for circumferential shell welds for two operating cycles is authorized pursuant to 10 CFR 7

50.55a(a)(3)(i).