JPN-98-022, Forwards Request for Approval of Alternative Plan,Iaw 10CFR50.55a(a)(3)(i),for RPV Shell Weld Examinations,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5).Summary of Commitments,Encl
| ML20248D346 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/28/1998 |
| From: | James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JPN-98-022, JPN-98-22, NUDOCS 9806020398 | |
| Download: ML20248D346 (12) | |
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- 123 Main Street White Plains, New York 10001 914 681 6840 914 287.3309 (FAX)
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May 28,1998 JPN-98-022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Sta; ion P1-137 Wathington, DC 20555
Subject:
James A FitzPatrick Nuclear Power Plant Docket No. 50-333 Proposed Alternatives in Accordance With 10CFR50.55a(a)(3)(i) for Reactor Pressure Vessel Shell Weld Examinations
Dear Sir:
Attached to this letter is a request for NRC approval of an alternative plan, in accordance with 10CFR50.55a(a)(3)(i), for the Reactor Pressure Vessel (RPV) shell weld examinations, pursuant to the provisions of 10CFR50.55a(g)(6)(ii)(A)(5).
The Authority will be parforming examination of the James A. FitzPatrick Nuclear Power Plant RPV shell welds during the next refueling outage (R13). The proposed examinations are an alternative to the augmented examinations for RPV shell welds specified 'in 10 CFR 50.55a(g)(6)(ii)(A), and an alternative to the inservice inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI,1989 (Table IWB-2500-1, Examination Category B-A, item No. B1.10.)
The proposed alternative plan for RPV shell welds includes examinations of the RPV vertical shell welds to the maximum extent practica! from the inner diameter within the constraints of vessel internal interference, and incidental examination coverage of 2 to 3 percent (14 to 22 inches) of the intersecting circumferential shell welds.
The remaining circumferential shell weld examinations would be deferred for two operating cycles per NRC Information Notice 97-63, Supplement 1. The anticipated extent of weld examination coverage for each vertical weld is identified in Table 1 of.
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. contains the Authority's request for approval and supporting justification for the alternative plan for the RPV shell weld examinations for the I
FitzPatrick plant. Review and approval of this alternative plan is requested prior to August 30,1998, to support planning of the outage scheduled to begin October 15, 1998.
Attachment il summarizes the commitments made by the Authority in this submittal.
If you have any questions please contact Mr. Art Zaremba at (315) 349-6365.
Very truly yours, o
J. Knubei jSenior Vice President and Chief Nuclear Officer cc:
Regional Administrator U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of the Resident inspector U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Joseph Williams, Project Manager Project Directomte 1-1 Division of Raactor Projects 1/11 U. S. Nuclear Regulatory Commission Mail Stop 14B2 Washington, DC 20555 Attachments 1.
Relief Request 15 - Proposed Alternative for Reactor Pressure Vessel Weld Shell Weld Examinations l
11.
Summary of Commitments Page 2
I Att:chment I to JPN 98-022 Relief Request 15 - Proposed Alternatives for Reactor Pressure Vessel Shell Weld Examinations Introduction l
The purpose of this letter is to request approval, pursuant to provisions contained in 10CFR50.55a(a)(3)(i), of an alternative plan for performing the reactor pressure vessel (RPV) augmented examination requirements of 10CFR55a(g)(ii)(A)(2) for the James A. FitzPatrick Nuclear Power Plant. The FitzPatrick alternative plan would require the performance of RPV verticel weld examinations from the inner diameter to the maximum extent possible within the constraints of vesselinternalinterference I
and the exam equipment's lower scan limitations. The alternative plan would only require incidental examination of the circumferential welds of 2-3 percent as result of vertical shell weld examinations and deferral of the remaining portions of the circumferential welds. The remaining portion would be deferred for two operating cycles (Fall 2002), pending NRC review, resolution, and approval of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) recommendations contained in "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," BWRVIP-05 (Reference 1). This is consistent with NRC information Notice 97-63, Supplement 1 (Reference 2).
The Authority endorses the current BWRVIP recommendations contained in BWRVIP-05. If the recommendations of the BWRVIP are changed duiing the approval process, the Authority will reevaluate the planned scope of examinations described in this attachmere in relation to conformance with the approved guideline.
A.
COMPONENT IDENTIFICATION ISI Class 1, Code Category B-A, " Pressure Retaining Welds in Reactor Vessel", item 81.10, "Shell Welds".
B.
EXAMINATION REQUIREMENTS 10CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implementing the examination requirements for Reactor Pressure Vesse: (RPV) shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWL of the 1989 Edition of Section XI, Division 1, of the ASME Boiler I
and Pressure Vessel Code, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). As stated in 10CFR50.55a, for the purposes of this augmented examination, essentially 100 percent as used in Table IWB-2500-1 meant more than 90 percent of the examination volume for each weld.
Additionally,10CFR50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the augmented RPV shell weld examination requirement to submit information to the U.S. Nuclear Regulatory Commission to support the determination, and propose an alternative to the examination requirements that would provide an acceptaMe level of quality and safety.
Page 1
Attahm:nt I to JPN-98-022 Relief Request 15 - Proposed Alternatives for
. Beactor Pressure Vessel Shell Weld Examinations C.
ALTERNATIVE TO THE EXAMINATION REQUIREMENTS The proposed examinations are an alternative to the augmented examinations for RPV shell welds specified in 10 CFR 50.55a(g)(6)(ii)(A)(2), and an alternative to the inservice inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI,1989 Edition (Table !WB-2500-1, Examination Category B-A, item No. B1.10). The proposed alternative plan for RPV shell welds includes examinations of the RPV vertical shell welds (Item B1.12) to the maximum extent practical from the inner diameter, within the constraints of vessel internal interference and examination equipment's lower scan limitations, and incidental examination of 2 to 3 percent of the intersecting circumferential shell welds (Item B1.11). The remaining accessible portions of the circumferential welds would be deferred for two operating cycles pending NRC review, resolution, and approval of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) recommendations contained in "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," BWRVIP-05. The deferral of the circumferential shell weld examinations for two operating cycles is consistent with NRC Information Notice 97-63, Supplement 1.
D.
BASIS FOR ALTERNATIVE PLAN The James A. FitzPatrick Nuclear Power Plant will be performing examinations of the RPV shell welds during the next refueling outage (R13). The Authority is unable to meet the 90 percent volume coverage requirement for each weld due to internal interferer ce of the FitzPatrick reactor vessel components (i.e., guide rods, shroud repair tie-rods, jet pump risers, feed water and core spray piping, etc.) and the examination equipment's lower scan limitations. The proposed alternative is to perform an examination of the RPV shell welds to the maximum extent practical from the inner diameter (lD), within the constraints of vessel internal interference.
1 Accessibility studies (Reference 3) of the FitzPatrick RPV have determined that the
- accessible area for volumetric examinations from the ID wi:1 allow coverage of approximately 60 percent of the cumulative length of the shell welds (vertical and circumferential welds). The anticipated extent of weld examination coverage for each vertical weld is identified on Table 1. Further examination from the ID is not practical without disassembly of vessel internal components. Removal of the RPV intamals would prebant undue hardship as a result of additional radiation exposure J
to personnel, increased costs, and increased outage duration, without a compensating increase in safety. Also, the BWRVIP-05 report provides additional L
. technical basis for only performing 50% of the RPV vertical shell welds. Only marginal increases in volumetric coverage could be obtained from additional outer I
diameter examinations. The additional OD examinations would result in increased dose of approximately 1.3 REM and an increased cost of over $300,000 due to
' direct labc.r without a compensating increase in safety.
l Page 2 1
Att:chmtnt I to JPN-98-022 Relief Request 15 - Proposed Alternatives for Reactor Pressure Vessel Snell Weld Examinations The alternative plan would require incidental examination of the circumferential welds of 2-3 percent at each examined vertical weld as result of vertical weld examinations and deferral of the remaining portions of the circumferential welds.
The remaining portion woukt be deterred for two operating cycles pending NRC review, resolution, and approval of the Boiling Water Reactor Vassel and Internals Project (BWRVIP) recommendations contained in "BWR Reactor Pressure Vessel Shell Weld inspection Recommendations,", BWRVIP-05. The basis for this request is
. documented in the BWRVIP-05 report that was transmitted to the NRC in September 1995. The BWP. VIP-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 for BWR RPV circumferential shell welds is several orders of magnitude lower than that of vertical shell welds (BWRVIP-05, section 8.3.2.4). The NRC staff has conducted an independent risk-informed assessment of the analysis contained in BWRVIP-05 (Reference 4). The NRC assessment also concluded that the probability of failure of the BWR RPV circumferential welds is orders of magnitude lower than that of the vertical shell welds. 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 conclusion of BWRVIP-05.
Previous Shell Weld Examinations During the fabrication process of the RPV, all of the shell welds were thorougnly examined using several examination methods as required by the original
. construction code. Additionally, all of the'shell welds received volumetric examinations prior to initial plant operations, as prescribed by ASME Section XI preservice inspection requirements. Selected shell welds have r' sived outer diameter volumetric examinations during the first and second in navice inspection interval in accordance with ASME Section XI inservice inspection requirements.
Only a few acceptable indications were identified during the first and second interval examinations. Table 1 shows the extent of the first and second interval weld examinations _ and the acceptable indications identified.
Cold Over-Pressurization 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 particular, the NRC stated that non-design basis cold over-pressure transients should be considered. It is highly unlikely that a BWR would
event would generally require several operator errors. At the meting of August 8, i
1997, the NRC described several types of events that could be precursors to BWR l'
_ Page 3 I
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Att:chment I ts JPN-98-022 Relief Request 15 - Proposed Altematives for Reactor Pressure Vessel Shell Weld Examinations RP_V cold over-pressure trJnsients. These were identified as precursors because no cold over-pressure event has occurred at a U.S. BWR. Also, at the August 8 meeting, the NRC identified one actual cold over-pressure event 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 Hiah Pressure Iniection Sources The high-pressure make-up systems at FitzPatrick (i.e., the Feedwater, High Pressure Coolant injection (HPCI), and the 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 plausible for these systems to contribute to an overpressurization event while the unit is in cold shutdown.
During reactor cold shutdown conditions the feedwater beaster pumps are normally maintained in the " pull-to-lock" position and the feedwater discharge isolation valves are normally maintained in the closed position. It would require several Operator errors and breakdowns in the work control process to inadvertently start a feedwater booster pump and inject into the vessel. As discussed below, operating procedural restrictions, operator training and work control processes at FitzPatrikc provide appropriate controls to minimize the potential for RPV cold over-pressurization events.
During normal cold shutdown conditions, RPV level and pressure are controlled with the Control Rod Drive (CRD) and Reactor Water Cleanup (RWCU) systems using a
" feed and bleed" process. The RPV is not taken solid during these items, and plant procedures require opening of the head vent valves after the reactor has been cooled to less than 212 F. If either of these systems were to fail, the Operator would adjust the other system 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 chayes and,' thus, significantly reduces the possibility of an event that would result in a viointn of the pressure-temperature limits.
]
The Standby Liquid Control (SLC) system is another high-pressure water source to the RPV. However, there are no automatic starts associated with this system. SLC injection requires an Operator 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 50 gpm, which would give the Operator ample time to control reactor pressure in the case of an inadvertent injection.
Pressure testing of the RPV is classified as an " Infrequently Performed Test or Evolution" which ensures that these tests receive special management oversight and procedural controls to maintain the plant's Svel of safety within acceptable limits.
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Att:chment I to JPN 98-022 Relief Request 15 - Proposed Alternatives for Reactor Pressure Vessel Shell Weld Examinations
-The pressure test is conducted so that the required temperature bands for the pressure increases are schieved and maintained prior to increasing pressure. During performance of an RPV pressure test, level and pressure are t,ontrolled using the CRD and RWCU systems using a " feed and bleed" process. Increase in pressure is limited to less than 30 psig per minute. This practice minimizes the likelihood of exceeding the pressure-temperature limits during performance of the test.
Procedural Controls /Ocerator Trainina to Prevent Reactor Pressure Vessel Cold Over-Pressurization 1
Operating procedural restrictions, operator training and work control processes at
~ FitzPatrick provide appropriate controls to minimize the potential for RPV cold over-pressurization events.
During normal cold shutdown conditions, reactor water level, pressure, and J
temperature are maintained within established bands in accordance with operating procedures. The Operations procedure governing Control Room activities requires that Control Room Operators frequently monitor for indications and alarms to detect abnormalities as early as possible. This procedure also requires that the Shift Manager be notified immediately of any changes or abnormalities in indications.
I Furthermore, changes that could affect reactor level, pressure, or temperature can only be performed under the knowledge and direction of the Shift Manager or Control Room Supervisor. Therefore, any deviations in reactor water level or 1
temperature from a specified band will be promptly identified and corrected. Finally, j
plant conditions and on-going activities that could affect critical plant parameters are discussed at each shift turnover. This ensures that on-coming Operators are cognizant of activities that could adversely affect reactor level, pressure, or
. temperature.-
Procedural controls for reactor tempemture, 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, Control Room Operators receive training on brittle fracture limits and compliance with the Technical Specification 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.
i During plant outages the work control processes ensures that the outage schedule
. and changes to the schedule receive a thorough shutdown risk assessment review i,
ito ensure defense-in-depth is maintained. At FitzPatrick, outage work items are scheduled by the Work Control Center. Senior Reactor Operators assigned to the Work Control Center provide oversight of outage schedule development to avoid conditions which could adversely impact reactor water level, pressure, or l.
temperature. From the' outage schedule, a daily schedule is developed listing the I
' work activities to be perforrned. These daily schedules are reviewed and approved Page 5 1
Att chment I to JPN-98-022 Relie).equest 15 - Proposed Alternatives for Reactor Pressure Vessel Shell Weld Examinations by Management, and a copy is maintained in the Control Room. Changes to the schedule require Management review and approval.
During outages, work is coordinated through the Work Control Center, which l
provides an additional level of Operations oversight. In the Control Room, the Shift Manager is required, by procedure, to maintain cognizance of any activity that could potentially affect reactor level or decay heat removal during refueling outages. The Control Room Operator is required to provide positive control of reactor water level and pressure within the specified bands, and promptly report when operating outside the specified band, including restoration actions being taken. Pre-job briefings are conducted for complex work activities, such as RPV pressure tests or hydrostatic testing that have the potential of affecting critical RPV parameters. Pre-job briefir.gs are attended by the cognizant individuals involved in the work activity.
Expeuud plant responses and contingency actions to address unexpected conditions, or responses that may be encountered, are included in the briefing
~ discussion.
Exam Eauioment The General Electric (GE) GERIS-2OOO System will be used to perform the remote controlled, automated UT examinations of the RPV. This tool has been used previously at other BWRs for the purpose of RPV examinations. GE demonstrated this system at the Performance Demonstration initiative (PDl), qualification session No. 61-02. The performance demonstrations were in acccdance with the PDI-RPV Protocol document implementing the requirements of Appendix Vlli of the 1992 edition with 1993 Addenda of the ASME Boiler and Pressure Vessel Code, Section
~ XI. Appendix Vill was developed to ensure the effectiveness of UT examinations within the nuclear industry by means of rigorous, item specific, performance demonstration. The performance demonstration was conducted on a RPV mockup containing flaws of various sizes and locations. The demonstration established the capability of the equipment, procedures, and personnel to find flaws that could be detrimental to the integrity of the RPV. Although Appendix Vill is not currently required by regulation, the qualification of equipment, procedures, and personnel to Appendix Vlli criteria demonstrates examination and evaluation techniques that surpass the requirements of the ASME Boiler and Pressure Vessel Code,Section XI,
. referenced by rule. General Electric's procedure is qualified per the PDI and is allowed per.lWA-2240 in lieu of the requirements contained in Table IWB-2500-1 of the ASME Boiler and Pressure Vessel Code,Section XI,1989 edition.
j
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Conclusion Based on the documentation in the BWRVIP-05 report, the risk-informed independent assessment performed by the NRC staff, and the discussion above, the i
Authority believes a deferralin completing the inspection of the RPV circumferential shell welds until the NRC reviews and approves the BWRVIP-05 recommendation is justified. As described in the industry basis document, BWRVIP-05; 1) the inherent Page 6
' Att:chmInt I ts JPN-98-022 Relief Request 15 - Proposed Alternatives for Reactor Pressure Vessel Shell Weld Examinations flaw tolerances of the boiling water reactor vessel due to lower radiation embrittlement and challenging design and operationalloadings,2) the quality of the
- original vessel fabrication,3) the lack of significant degradation mechanisms, and 4) the results of previous vessel examinations, provide the basis for concluding that the proposed alternative plan to perform extensive and distributed, high-quality, vessel shell weld examinations will provide an acceptable level of quality and safety.
The Authority will be working with the BWRVIP to resolve the longer-term issues in this area, but believes BWRVIP-05 and the NRC analysi provide sufficient basis to support this alternative plan request.
. E.
ALTERNATIVE EXAMINATIONS The RPV vertical shell welds will be volumetrically examined to the maximum extent practical from the inner diameter within the constraints of vesselinternal interference. The RPV circumferential shell welds will have incidental volumetric examinations at intersecting points of each examined vertical weld to an extent of 2.
to 3 percent (14 to 22 inches) as result of vertical weld examinations. The remaining portion of the RPV circumferential shell welds will be deferred for two operating cycles (Fall 2002) pending NRC review, resolution, and approval of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) recommendations contained in the BWRVIP-05 report. If the recommendations of the BWRVIP are changed during the approval process, the Authority will reevaluate the planned scope of examinations described in this attachment in relation to conformance with the approved guideline.
The Authority will provide a summary report of the results of the RPV examinations to the NRC no later then 90 days from the completion of the outage, required by Article IWA-6000 of the ASME Boiler and Pressure' Vessel Code,Section XI. The submittal will include the individual weld coverage obtained with the GERIS-2OOO System, and the examination results if relevant indications are found during the examination, an engineering acceptance flaw evaluation will be performed prior to startup from the refueling outage.
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AttIchment I to JPN-98-022 Relief Request 15 - Proposed Alternatives for Reactor Pressure Vessel Shell Weld Examinations References
- 1. EPRI TR-105697, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), September 1995.
- 2. NRC Information Notice 97-63, Supplement 1, " Status of NRC Staff's Review of l
BWRVIP-05," dated May 7,1998.
- 3. GE Nuclear Energy Accessibility Study, GENE B13-01869-081, Revision 0, June I
1997.
- 4. NRC letter, Brian W. Sheron (NRC) to Carl Terry (BWRVIP Chairman),
" Transmittal of NRC Staff's Independent Assessment of the Boiling Water Reactor Vessel and Internals Project BWRVIP-05 Report and Proprietary Request for Additional Information (TAC No. M93925)," dated August 14,1997.
"B5NRVIP Response to NRC Rcquest for Additional Information on BWRVIP-05,"
dated December 18,1997.
- 6. GE Nuclear Energy, GERIS 2000 ID System Alternate Method of Volumetric Examination, February 1998.
I i
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I I
i Page 9
AttIchm:nt il to JPN-98-022 I,
Relief Request 15 - Proposed Alternatives for l
Reactor Pressure Vessel Shell Weld Examinations J
l l
i Summary of Commitments I
Commitment Number Description Due Date JPN-98-022-OU1 Provide a summary report No later then 90 days of the results of the RPV from the completion of the examinations to the NRC.
outage, required by Article The submittal will include IWA-6000 of the ASME the individual weld Boiler and Pressure Vessel coverage obtained with Code,Section XI.
the GERIS-2000 System, and the examination results.
JPN-98-022-002 If relevant indications are Prior to startup from the found during the refueling outage.
examination, an engineering flaw acceptance evaluation will be performed.
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