ML20199F554
| ML20199F554 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/17/1997 |
| From: | Jury K CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9711240259 | |
| Download: ML20199F554 (10) | |
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- Caronna Powe, & IJght Company 7 PO Box _10429.
Soutevt, NC -284810429 -
LNovemb:r 17,1997' SERIAli USEP 97-0489 -
10 CFR 50.55a(a)(3) 10 CFR 50M5a(g)(6)(ii)(A)(5)
U. S. Nuclear Regulatory Commission.
ATTN: Document Control Desk Washington, DC 20555 -
BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50-325/ LICENSE NO. DPR-7 i REQUEST FOR APPROVAL OF ALTERNATIVE TO INSPECTION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELDS Gentlemen:
In accordance with 10 CFR 50.55a(a)(3), Carolina Power & Light (CP&L) Company requests approval of an ah a sative reactor vessel weld examination for the f3runswick Steam Electric Plant (BSEP), Unit No.1. Approval of this alternative examination is requested in accordance with the provisions of 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(5) for the next two 4
operating cycles.
CP&L proposes to perform ins),ections of essentially 100 percent of the longitudinal seam welds in the reacter pressure vessel (RPV) shell and essentially 0 percent of the RPV circumferential seam welds, which will result in partial examination (i.e., approximately 2 to 3 percent) of the circumferential welds at their points of'atersection wi*h the longitudinal welds. These inspections are being proposed as an attemative to the augmented examinations specified in 10 CFR 50.55a(g)(6)(ii)(A)(2) for circumferential welds, and as an alternative to the inservice inspection requirements for circumferential welds in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI,1980 Edition with Winter 1981
- Addenda.' Further discussion of the basis for this request is enclosed. No new regulatory commitments are contained in this letter. A.similar request for BSEP, Unit No. 2 was approved by an NRC letter dated September 18,1997.
p i
. BSEP, Unit No.1 Refueling Outage 11 (i.e., Bl12RI) is scheduled to begin on April 25,1998.
1 in order to supprt planning activities for the necessary inservice inspections, approval of this j
proposed alternative for BSEP, Unit No. i is requested by March 25,1998.
9711240259'.971117
- PDft ADOCK 05000325 I-N
y
~ Document Control Desk
- - BSEP 97-0489 / Page 2_.:
- Please refer any questions regarding this submittal to Mr. Warren J. Dorman, Supervisor - -
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_ Licensing at (910)457-2068.
Sincerelyi Keith R. Jury Manager-Regulatory Affairs Brunswick Steam Electric Plant
-:WRM/wrm
Enclosure:
Request For Approval of Alternative to inspection of Reactor Pressure Vessel Circumferential Welds cc (with enclosure):
- U. S. Nuclear Regulatory Commission, Region 11 ATTN: Mr. Luis' A. Reyes, Regional Administrator Atlanta Federal Center -
i 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 U. S. Nuclear Regulatory Commission ATTN: Mr. Charles A. Patterson, NRC Senior Resident inspector 8470 River Road Southport, NC 28461 -
U. S. Nuclear Regulatory Commission
- ATTN: Mr. David C.Trimble, Jr. (Mail Stop OWFN 141122) 11555 Rockville Pike Rockville, MD 20852 2738 The lionorable Jo A. Sanford Chairman - North Carolina Utilities Commission -
PO Box 29510 Raleigh, NC 27626-0510
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Document Control Desk
- IISEP 97 0489 / Page 3 Division of Boiler and Pressure Vessel North Carolina Department of1. abor
~ ATTN: Mr. Jack Given, Assistant Director of Boiler & Pressure Vessels 4 West Edenton Street -
Raleigh, NC 27601-1092 i
4'.
ENCLOSURE
. BRUNSWICK 3 TEAM ELECTRIC PLANT, UNIT NO. I DOCKET NO. 50-325/ LICENSE NO. DPR-71 REQUEST FOR APPROVAL OF ALTERNATIVE TO INSPECTION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELDS INTRODUCTION In accordance with 10 CFR 50.55a(a)(3), Carolina Power & Light (CP&L) Company requests approval of an alternative reactor "essel weld examination for the Brunswick Steam Electric Plant (BSEP), Unit No.1. Approval of this alternative examination is requested in accordance with the provisions of 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(5) for the next two operating cycles. The alternative is consistent with information contained in NRL Information
- Notice 97-63, " Status of NRC Staff Review of BWRVIP-05."
The alternative being proposed is the performance ofinspections of essentially 100 percent of the BSEP, Unit No. I reactor pressure vessel (RPV) shell longitudinal seam welds and essentially 0 percent of the RPV shell circumferential seam welds during Refueling Outage i1 (i.e.,
B112Rl), which will result in partial examination (i.e., approximately 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 Specification 4.0.5 for BSEP, Unit No.1. Technical Specification 4.0.5 requires that the inservice inspection and testing of the American Society of hiechanical Engineers (ash 1E) Code Class 1,2, and 3 components be perfomied in accordance with Section XI of the ash 1E Boiler and Pressure Vessel Code, and applicable addenda, as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i),10 CFR 50.55a(a)(3) allows alternatives to the requirements of paragraph (g) to be used, when authorized by the NRC if the proposed alternative would provide an acceptable level of quality and safety, in accordance with 10 CFR 50.55a(g)(4), AShlE Code Class 1,2, and 3 components must meet the requirements, except the design and access provisions and the presersice examination requirements, in the ash 1E 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 construcuan of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of the ASME Code,Section XI, incorponted by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 10-year interval, subject to the limitations and modifications listed therein. The applicable ASNIE Code,Section XI, for BSEP, Unit No. I during the second 10 year inservice inspection interval is the 1980 Edition through the Winter 1981 Addenda. The __
l com;ionents, including supporta, may meet the requirements in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modi 6 cations listed, ur.d subject to NRC approval.
10 CFR 50.55a(g)(6)(ii)(A) requires the performance of an expanded RPV shell weld examination, as specined in the 1989 Edition of the ASME Code,Section XI, on an " expedited" basis. " Expedited," in this context, effectively meant during the inspection interval when the regulation was approved or the Drst period of the next inspxtion interval The regulation was published in the Federal Register (57 FR 34666) on Au"ust 6,1992. By incorporating into the regulations the 1989 Edition of the ASME Code, the NRC required performance of volumetric examinations of" essentially 100 percent" of the RPV pressure retaining shell welds during all inspection intervals.
On May 12,1997, the NRC and members of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) met to discuss the NRC's review of the BWRVIP-05 report entitled "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations." In accordance with guidance provided in NRC Staff Requirements Memorandum M970512B, dated May 30,1997, the NRC initiated a broader, risk-informed review of the proposal contained within BWRVIP-05.
In Infonnation Notice 97-63, the NRC indicated that it would consider technicallyjustified alternatives to the augmented examination, in accordance with 10 CFR 50.55a(a)(3)(i),
10 CFR 50.55a(a)(3)(ii), and 10 CFR 50.55a(g)(6)(ii)(A)(5), from boiling water reactor licensees who are scheduled to perform inspections of the RPV shell circumferential welds during the Fall 1997 or Spring 1998 outage periods. The Information Notice indicated that acceptably justified alternatives would be considered for inspection delays of up to 40 months, or two operating cycles, whichever is longer, for RPV circumferential shell welds only.
JUSTIFICATION Reactor Vessel Integrity:
The basis for CP&L's request is documented in the BWRVIP-05 report, which was transmitted to the NRC in September 1995. The BWRVIP-05 report provides the technical basis for eliminating inspection of BWR RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitade lower than that of the axial shell welds. The NRC conducted an independent risk-informed assessment of the analysis contained within the BWRVIP-05 document. The BWRVIP-05 report indicates that, for r. 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. Additionally, the NRC assessment demonstrated that inspection of BWR RPV circumferential welds does not measurably affect the probability of failure. Therefore, the NRC evaluation appears to support the conclusions of the BWRVIP-05 report.
This independent NRC assessment used the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the l
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PFM nnalysis are: the neutron fluence was that estimated to be end of-liccuse mean fluence, the chemistry valuer sre mean values based on vessel types, and the potential for beyond design basis events is censidered.
Although the llWRVIP-05 report provides the technical basis supporting the alternative examination sequest, the following infonnation is provided to show the conservatisms of the NRC analy sis lbr llSEP, Unit No.1. Changes la RTu:,r may be used as one of the menns for monitoring radiation embrittlement of reactor vessel materials. In the case ofIIShP, Unit No.1, the single circumferential weld joint located within the RPV beltline would be the limiting circumferential weld within the vessel (i.e., relative to RTum). For plants with RPVs fabricated by Chicago liridge & Iron (Cl3&l), the mean end-of license neutron lluence used in the NRC 2
i PFM analysis was 0.19E+19 n/cm. Ilowever, the highest fluence anticipt ted at the end of the 2
next two operating cycles is 0.063E+19 n/cm for 11SEP, Unit No.1. hs, the fluence efTect on embrittlement is much lower, and the NRC analysis described at an August 8,1997, meeting with industry is conservative for llSEP, Unit No.1 in this regard. Therefore, there is significant
. conservatism in the already inw circumferential weld failure probabilities as related ta llSEP, Unit No.1. Other llSEP Unit No.1 RPV shell weld infbnnation is included in Table 1 of this letter.
L As shown in Table 1, the calculated embr;ttlement shift in RTsm (i.e., ARTsm) for the llSEP, Unit No. n vessel is 27.13 F at the end of the requested period. Ily comparison, using the mean 2
values fbr iluence (i.e.,0.191 n/cm ) and weld chemistry (i.e., copper content of 0.04 weight percent, nickel content of 0.93 webiht percent) assumed for Cll&l reactor vessels in Table 7-5 of Enclosure I to the NRC independent assessment report, a ARTsor f 30.16 F would be derived.
o Therefore, the calculated ARTum value Ihr the llSEP, Unit No. I vessel is less than, and thas bounded by, the embrittlement shill assumed in the NRC's independent assessment.
Furthennore, as seen in the attached Table 1, the calculated Upper llound RTum value fbr the llSEP, Unit No. I beltline welds is 64.26' F at the end of the requested period, based on the initial RTum values fbr these weld joints provided in CP&L's letter dated November 16,1995 (Serial: ilSF.P 95 0572;. Using the generic value of 56 F for initial RTum yields an Upper 14 und Rl'um value of 14.63* F for 13SEP, Unit No.1. For comparison, the highest Upper llound 0
RTum value [i.e., " Inner Suiface (RTum + 2a) F"] shown within Tables 7 6,7 7, and 7-8 of to the NRC's independent assessment report of the llWRVIP-05 document, would be the RTum of 145.l* F shown within Table 7-7 for the 11&W fabricated ilWR vessels. Again, the calculated Upper Bound RTum values for the llSEP, Unit No.1 vessel circumferential welds are clearly bounded by the limiting RTum from Table 7 7 of the NRC independent assessment report, thus providing addition:d assurance that the llSEP, Unit No.1 vessel welds are also bounded by the llWRVIP 05 report..
Cold Over Pressurizati D:
9 At the meeting on August 8,1997, the NRC indicated that the potential for, and consequences of, non design basis events not addressed in the BWRVIP-05 report should be considered. in particelar, tne NRC stated that non design basis cold over-pressure transients should be e
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considered it is highly unlikely tnat a 11WR would experience a cold over pressure transient, in a
fact, for a IlWR te experience such an event would generally require several operator errors. At the meeting of August 8,1997, the NRC described several types of events that could be precursors to BWR RPV cold over-pressure transients. These were identified as precursors because no cold over pressure event has occurred at a U. S.11WR. Also, at the August 8 l
meeting, the NRC identified one actual cold over pressure ewat that occurred during shutdown at a non U. S. BWR. This event apparently included several operational errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79"F to 88'F.
Review of Potential.Ulch Pressure Inlection Sources:
The high pressure make up systems at BSEP [i.e., the liigh Pressure Coolant Injection (llPCI) and Reactor Core isolation Cooling (RCIC) systems] are steam turbine driven. During reactor cold shutdown conditions, no steam is available for operation of these systems. Therefore, it is not plataible for these ystems to contribute to an overpressurization event while the unit is in cold shutdown.
Two precursor events are identified for IlSEP Unit No. 2 in Table C 1 of the NRC's Independent
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Assessment Report for BWRVIP 05; no precursor events for BSEP, Unit No. I are identified in this NRC report. Both of the BSEP, Unit No. 2 events involved inadvertent injectie oflow pressure makeup systems. The first event resulted in an injection of the Low Pressure Coolant injection (LPCI) and Core Spray systems, and the second event resulted in an injection of the Core Spray system. Neither of these events resulted in a violation of the pressure temperature limits. The llSEP pressure-temperature limit curves for hydrostatic testing allow pressures up to 313 psig at a temperatur of 70* F. The shutoff heads for the BSEP Core Spray and LPCI system pumps are approximately 313 psig and 250 psig, respectively. Therefore, the potential for an overpressurinition event which would exceed the pressure / temperature limits, due to the inadvertent actuation of these systems, is very low.
During normal cold shudown conditions, RPV level and pressure are controlled with the Control Rod Drive (CRD) cud Reactor Water Cleanup (RWCU) systems using a " feed and bleed" process. The RPV is not taken solid during these times, and plant procedures require opening of the head vent valves atter the reactor has been cooled to less than 212 F. If either of these systems were to fail, the Opeiator would adjust the other systeni to control level. Under these conditions, the CRD system typically injects water into the reactor at a rate of <60 gpm.
This slow injection rate allows the operator sufficient time to react to unanticipated level chaages ard, thus, significantly reduces the possibility of an event that would result in a violation of the i
pressure temperature limits.
The Standby Liquid Control (SLC) system is another high pressure water source to the RPV.
Ilowever, there are no automatic starts associated with this system. SLC injection requires an Oocrator to manually start the system from the Control Room or from the local test station.
Additionally, the injection rate of the SLC pump is approximately 41 gpm, which would gise the Operator ample time to control reactor pressure in the case of an inadvertent injection. -.
4 Pressure :esting of the RPV is classified as an " Infrequently Performed Test or Evolution," which ensures that these tests receive special management os ersight and procedural controls to maintain the plant's leve! of safety within acceptable limits. IISEP practice is to heat up the reactor to hydrostatic test temperature (approximately 170 to 180' F) prior to increasing pressure. During performance of an RPV pressure test, level and pressure are controlled with the CRD and RWCU systems using a " feed and bleed" process. Increr.se in pressure is limited to 50 psi per minute.
This practice minimizes the likelihood of exceeding the pressure temperature limits during perfonnance of the test. In addition, the operating procedures used for 11SEP, Unit No. I require that the safety relief valves be installed and operable during the performance ofleak tests performed at the conclusion of each outage.
Procedural Controls /Onerator Trainine to Prevent Reactor Pressure Vessel Cold Over-Pressurization:
Operating procedural restrictions, operator training, and work control processes at ilSEP provide appropriate controls to minimize the potential for RPV cold over pressurization events.
During normal cold shutdown conditions, reactor water level, pressure, and temperature are maintained within established bands in accordance with operating procedures. The Operations procedure governing Control Room activities requires that Control Operators (COs) frequently monitor for indications and alarms to detect abnonnalities as early as possible. This procedure also requires that the Senior Control Operator (SCO) be notified immediately of any changes or abnonnalities in indications. Furthennore, this procedure requires that changes which could affect reactor level, pressure, or temperature only be perfonned under the knowledge and direction of the SCO. Therefore, any deviations in reactor water level or temperature from a specified band will be pmmptly identied and corrected. Finally, the status of plant conditions, any on going activities which could affect critical plar.t parameters, and contingency planning are discussed by Operators at each shill turnover. This ensures that on-coming Operators are cognizant of any activiiies which could adversely affect reactor level, pressure, or temperature.
A review ofindustry operating experience indicates that inadequate work management is a potential contributor to a c )ld over pressurization event. At ilSEP, work performed during outages is scheduled by the Outage Management group. Dedicated Senior Reactor Operators provide oversight of outage schedule development to avoid conditions which could adversely impact reactor water level, pressure, or temperature. From the outage schedule, a plan of the day (POD)is developed listing the work activities to be performed. These PODS are reviewed and approved by Management, and a copy is maintained in the Control Room. Changes to the PODS require Management review and approval. Additionally, the detailed outage schedule receives a thorough shutdown risk assessment review to ensure defense in-depth is maintained.
During outages, work is coordinated through the Work Control Center, which provides an additional level of Operations oversight. In the Control Room, the SCO ir required, by procedure, to maintain cognizance of any activity which could potentially affect reactor level or decay het removal during refueling outages. The CO is required to provide positive control of reactor water level and pressure within the specified bands, and promptly report when operating _,. _..
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' outsiile the specified band, including restoration actions being taken. Pre job briefings are conducted for work activities that have the potential of affecting critical RPV parameters. These briefings are af tended by the cognizant individuals involved in the work activity. Expected plant responses and contingency actions to address unexpected conditions or responses that may be encountered are included in the briefing discussion.
The plant procedure for unit shutdown limits reactor pressure to $10 psig while flooding up to cold shutdown ster level and requires frequent monitoring of reactor pressure to ensure that this limit is not exceeded. Additionally, this procedure requires opening of the head vent valves afler the reactor has been cooled to less than 212' F. Procedural controls for reactor temperature, level, and pressure are an integral part of Operator training. Specifically, Operators are trained in methods of controlling water level within specified limits, as well as responding to abnormal water level conditions outside the established limits. Additionally, COs receive training on brittle fracture limits and compliance with the Technical Specificatien pressure temperature limits curves. Plant-specific procedures have been developed to provide guidance to the Operators regarding compliance with the Technical Specification requirements on pressure-temperature limits. In addition to the existing training, on August 28,1997, a memmandum was transmitted from Engineering to Operations to further raise awareness of the potential for cold over pressurization, citing the available informsion on industry events.
SUNih1ARY:
The BWRVIP-05 report provides the technical basis for elim:nating inspection of BWR RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the llWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. Ilased on an assessment of the materials in the circumferential weld in the beltline of the BSEP, Unit No. I RPV, the conditional probability of RPV failure should be less than or equal to that estimated in the NRC's independent assessment. Based on Operator training and established procedures, the probability of cold over-pressure transients should be minimized during the next two operating cycles. Therefore, the probability of a cold over pressure transient is considered to be less than or equal to that used in the NRC. ssessment described at the meeting on August 8,1997, and is conservative for BSEP, Unit No.1.
The NRC staff has recently transmitted a Request for Additionai information (RAI) regarding the BWRVIP 05 report to the BWR Vessel and Internals Project (BWRVIP) The BWLVIP plans to provide a response to that RAI in the near future that will include additional informa ion on the BWRVIP probabilistic failure mode (PFht) analysis, comparisons to the NRC staff PFht analysis ano additional information regardmg beyond design basis cold over pressure transients. CP&L will work with the BWRVIP to resolve the longer term issues in this area, but CP&L believes the llWRVIP-05 report and the NRC analysis provide sufticient basis to support approval of this alternative examination request.
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o TABLE 1 ilSEP, Unit No. I RPV Shell Weld Infonnation 1
Neutron fluence at the end of the next two 0.063E+19 n/em' operating cycles Initial (unirradiated) reference temperature Value estimated by Method 4 of
+10'F (Estimated Value)
.l e
MTEli 5 2 and submitted in CP&L's Itesponse dated November 16,1995, to Generie Letter 92-01 NRC generic value
-56*F (Generie Value)
+
I Weld chemistry factor (CF) 82.0" F Weld copper content 0.06 %
increase in reference temperature due to 27.13' F irradiation (ARTwm)
Margin term 27,13' F C' 43.50" F
- Mean adjusted reference temperature 37.13* F "'
(Mean ART)
Upper bound adjusted referenec 64.26'F m temperature (ART) 14.63'F *
- Notes:
1, Value based on usage of an initial RTum of +10 F (estimated using MTED 5 2, Estimatior
~ Method 4).
- 2. Value based on usage of an initial RTum of-56'F (generie value).
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