ML20216B104
| ML20216B104 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/04/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216B103 | List: |
| References | |
| NUDOCS 9803120412 | |
| Download: ML20216B104 (5) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCl FAR REACTOR REGULATION
' ALTERNATIVE TO INSPECTION OF REACTOR PRESSURE VESSEL CIRCUMFERE.NTIAL WELDS BRUNSWICK STEAM FI FCTRIC PLANT UNIT 1 l
CAROLINA POWER & LIGHT COMPANY DOCKET NO: 50-325 1.0 Introduction 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 attemative to performing the reactor pressure vessel (RPV) circumferential shell wild examinations requirements of both the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel 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 attemative was proposed pursuant to the provisions of 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(ii)(A)(5), and is consistent with information contained in Information Notice (IN) 97-63,
" Status of NRC Staff Review of BWRVIP-05."
The attemative 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 zero percent of the RPV shell circumferential seam welds during Refueling Outage 11, which will result in partial examination (2 to 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 Technical Specifications (TS) 4.0.5 for BSEP Unit 1 which state that "the inservice inspection (ISI) and testing of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components be performed in accordance with Section XI of the ASME Boiler j
and Pressure Vessel Code, and applicable addenda, as required by 10 CFR 50.55a(g), except where spacific 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 design and access provisions and the preservice examination requirements, set i
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 regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and Attachment 9003120412 900304 PDR ADOCK 05000325 G
1 i subsequent intervals comply with the requirements in the latest edition and addenda of the ASME Code,Section XI, incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable ASME Code,Section XI, for BSEP 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 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 penod of the next inspection interval. The final Rule was published in the Esdera/ Registeron August 6,1992 (57 FR 34666). By incorporating into the regulations the irs 89 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(s)(3)(i) (10 CFR 50.55a(a)(3)(i)) indicates that attematives to the requirements in 10 CFR 50.55a(g) are justified when the proposed altamative 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, June 13, and December 18,1997, the Boiling Water Reactor Vessel and Intemals Project (BWRVIP), a technical committee of the BWR Owners Group (BWROG), submitted the proprietary report, "BWR Vessel and Intemals Project, BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," which proposed to reduce thc scope ofinspection 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 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 to 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 staffs review of the BWRVIP-05 report. In accordance with guidance provided by the Commission in Staff Requirements Memorandum (SRM) M9'/0512B, 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 attematives to the augmented examination in accordance with 10 CFR 50.55a(s)(3)(i),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 attematives would be considered for inspection delays of up to 40-months i
or two operating cycles (whichever is longer) for BWR RPV circumferential shell welds only.
2.0
. Background - Staff Assesunent of BWRVIP-05 Report The staffs 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
-h..
, industry's assessment does not sufficiently address risk, and additional work is necessary to provide a complete risk-informed evaluation.
The sta#s assessment was performed for BWR RPVs fabricated by Chicago Bridge and Iron (C3&I), Combustion Engineering (CE), and Babcock & Wilcox (B&W). The staff assessment idt 1tified 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 sta#s independent assessment, the staff determin6d the conditional probability of failure for axial and circumferential welds fabricated by CB&l, CE, and B&W. Table 7-9 of the sts#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&l, 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 (RTm7) 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 RTer values less than those resulting from the stars assessment will have less embrittlement than the vessels simulated in the sta#s assessment and should have a conditional probability of vessel failure less than or equal to the values in the sta#s assessment.
The failure probability for a weid 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 probability of vessel failure for CB&l fabricated circumferential welds, the best-estimate failure frequency from the sta#s assessment is
<6.0 X 10 u m per reactor year and the uncertainty bound failure frequency is <2.8 X 10mm 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 attemative 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 J
welds. This conclusion was also demonstrated in the stars independent assessment of the BWRVIP-05 report. The BWRVIP-05 report indicates that, for a typical BWR RPV, the failure i
probability for axial welds is 2.7 X 10 and the failure probability *or circumferential welds is 2.2 X 10* for 40 years of plant operation.
The licensee calculated the RT, value for the limiting BSEP Unit 1 circumferential weld at the end of the requested relief period using the methodology in Regulatory Guide (RG) 1.99, i
Revision 2. The RTer values calculated in accordance with RG 1.99, Revision 2, depend upon j
the neutron fluence, the amounts of copper and nickel in the circumferential weld, and its unirradiated RTer. The licensee determined the highest neutron fluence at the end of the next two operating cycles at the inner surface of the limiting circumferential beltline weld to be 1
Insufficient or no failures to accurately determine reference case failure probability.
s I
4 0.063 X 10" n/cm. The amounts of copper and nickelin the limiting circumferential beltline I
2 weld is 0.06 percent and 0.87 percent, respectively. The plant specific unirradiated RT for the limiting 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 RT resulted from the plant specific unirradiated RT value that was provided in CP&L's letter dated November 16,1995, and that use of the generic initial RTavalue of-56*F would yield an RT value of 14.63*F. The larger RT value that results from the plant specific unirradiated RT is still less than the most limiting reference case (B&W fabricated vessels) in the stats assessment. Since the RT of the BSEP Unit 1 beltline circumferential weld is less than the limiting RTa value in the staffs independent assessment, the licensee concluded that the BSEP Unit 1 vessel circumferential welds are bounded by the f.taft's independent assessment, thus providing additional assurance that the vessel welds are also bounded by the BWRVIP-05 report.
The licensee assessed the systems that could lead to a cold over-pressurization of the BSEP -
Unit 1 RPV. These included the high pressure coolant injection (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. The licensee stated that there were no automatic starts associated with SLCS. Operator initiation of SLCS should not occur during shutdown, however, the SLCS injection rate is approximately '1 gpm which would allow operators sufficient time to control reactor pressure if manual initiation occurred. The CRD and RWCU systems use 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 unanticipated level changes.
In all cases, the operators are train 4 in methods of controlling water level within specified limits in addition to responding to abnorn,al 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 i
over-pressure transient is unlikely to occur during the requested delay. 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 staffs assessrnent of BWRVIP-05.
4.0 Staff Review of Ucensee Technical Justifcation BSEP Unit 1 is a CB&l fabricated vessel, and the staff noted that the RTa value determined from the plant specific unirradiated RTa(64.26'F) is approximately 14*F higher than the limiting value determined in the staffs assessment for CB&l fabricated vessels (50*F).
However, RT is a measure of the amount of irradiation embrittlement, and since CB&l fabricated vessels have very low copper values, they have low amounts of irradiation embrittlement. For comparison, the staff confirmed that the RT value for the circumferential welds at the end of the relief period is less than the values in the limiting reference case and
I uncertainty analysis for the B&W fabricated vessels. Since the RT, values are well below the values in the reference case and the uncertainty analysis for B&W fabricated vessels, the BSEP Unit 1 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 staffs 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 over-pressurization. The information provided sufficient basis to support approval of the alternative examination request. Based on the high pressure injection systems analyses, operator training, and plant specific procedures, the probability of a cold over-pressurization transient occurring at BSEP Unit 1 during the requested delay is low, which is consistent with the staffs assessment.
5.0 Conclusions I
1 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 Brunswick Unit 1 RPV, the conditional probability of vessel failure should be less than or equal to that estimated from the staffs assessment of the :lmiting case of B&W fabricated vessels.
2)
Based on the licensee's high pressure injection systems analyses, operator training, and j
plant specific procedures the probability of cold over-pressure trar tients should be sufficiently low during the requested delay period.
3)
Based on the previous two conclusions, the staff concludes that the Brunswick Unit 1 l
RPV can be operated during the requested delay period with an acceptable level of quality and safety and the inspection of the circumferential welds may be delayed for the requested two operating periods.
Therefore, the proposed attemative to performing the RPV examination requirements of the ASME Code,Section XI,1980 Edition, through 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 50.55a(a)(3)(i).
l.