JPN-99-025, Forwards Relief Request 17 Re Proposed Alternatives IAW 10CFR50.55a(a)(3)(i) for Reactor Pressure Vessel Circumferential Shell Weld Exams

Forwards Relief Request 17 Re Proposed Alternatives IAW 10CFR50.55a(a)(3)(i) for Reactor Pressure Vessel Circumferential Shell Weld Exams

ML20210M806
Person / Time
Site:	FitzPatrick
Issue date:	08/05/1999
From:	James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-98-05, JPN-99-025, NUDOCS 9908100184
Download: ML20210M806 (12)

Text

{{#Wiki_filter:* 123 Main Street White Plains, New York 10S01 914 681.6950 914 287.3309 (Fax) A NmW&%er a - xa e i tv Authority sW1030"l*"" August 5,1999 JPN-99-025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, DC 20555

Subject:

James A FitzPatrick Nuclear Power Plant (Relief Request #17) Docket No. 50-333 Proposed Alternatives in Accordance with 10CFR50.55a(a)(3)(1) for Reactor Pressure Vessel Circumferential Shell Weld Examinations

Reference:

NRC Generic Letter 98-05, " Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," dated November 10,1998.

Dear Sir:

This letter transmits Relief Request #17 to the JAF ISI Program for permanent (i.e., for the remaining term of operation under the existing, initial, license) relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential reactor pressure vessel (RPV) welds. NRC Generic Letter 98-05 (Reference) states that "BWR Licensees may request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential reactor pressure vessel welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item B1.11, Circumferential Shell Welds) by demonstrating that: (1) at the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staff's July 30,1998, safety evaluation, . and (2) licensees have implemented operator training and established procedures that limit the frequency of cold-over-pressure events to the amount specified in the staff's July 30, 1998, safety evaluation." Attachment 1 contains the Authority's supporting justification and basis for the permanent relief from the RPV circumferential shell weld examination requirements for the FitzPatrick plant. Similar relief has been granted for other BWRs, such as Nine Mile Point Unit 1which has a Combustion Engineering fabricated vessel similar to FitzPatrick NPP. f 9908100184 990005 PDR ADOCK 05000333 P PDR Y

.i :, If.you have any questions, please contact Mr. George Tasick at (315) 349-6572. Very truly yours, nubel t Senior Vice President and Chief Nuclear Officer cc: Regional Administrator U. S. Nuclear Regulato.y Commission 475 Allendale Road King of Prussia, PA ',9406 Office of the Resident inspector U. S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. G. Vissing, Project Manager Proie t Directorate i Division of Licensing Project Management U. S. Nuclear Regulatory Commission Mail Stop 8C2 Washington, DC 20555 i k h t .I b): i _. . 'g ;

Att chm:nt i JPN-99-025 Page 1 of 7 ~ p' Relief Request 17 j Relief Request from ASME Section XI Code Regarding Reactor Vessel Circumferential Shell Weld Examinations A. COMPONENT IDENTIFICATION: ISI Class 1, Code Category B-A, " Pressure Retaining Welds in Reactor Vessel", item B1.11, "Circumferential Shell Welds". RPV Circumferential Welds: VC-1-2, VC-2-3, VC-3-4, VC-4-BH-1. (Attachment 11 contains a drawing which depicts these components.) 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 Vessel (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 lWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). As stated in 10CFR50.55a(g)(6)(ii)(A)(2) for the purposes of this augmented examination, essentially 100 percent as used in Table IWB-2500-1 means 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 acceptable level of quality and safety. C. ALTERNATIVE TO THE EXAMINATION REQUIREMENTS The alternative plan would require performance of RPV vertical weld examinations and incidental examination of 2 to 3 percent of the intersecting circumferential shell welds to the maximum extent possible based on accessibility. The circumferential welds would be permanently defw.d until plant operating license expiration. The proposed deferral is an alternative to the augmented examinations for RPV shell welds specified in 10 CFR 50.55afg)(6)(ii)(A)(2) and is consistent with NRC Generic Letter 98-05 (Reference 1). D. BASIS FOR ALTERNATIVE PLAN: On November 10,1998, the NRC issued Generic Letter 98-05 (GL 98-05), " Boiling Water Reactor Licensees Use of the BWRVIP-05 Report (Reference 2) to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds." The Generic Letter provided guidance for licensees seeking relief from requirements for examination of BWR RPV circumferential shell welds as recommended in the BWRVIP-05 report. GL 98-05 states that the licensee may request relief from the in-i service inspection requirements of 10CFR50.55a(g) for volumetric examination of the RPV circumferential shell welds by demonstrating: (1) at the expiration of their license, the circumferential shell welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staff's July 30,1998, safety evaluation, and (2) licensees

Attachm nt i JPN-99-025 Page 2 of 7 have implemented operator training and established procedures that limit the frequency of cold over pressure events to the amount specified in the staff's July 30,1998, safety evaluation. 1.- The following response provides the basis that demonstrates that at the expiration of the James A. FitzPatrick NPP operating license the FitzPatrick circumferential shell welds will continue to satisfy the limiting conditional failure probability for circumferential welds stated in the staff's July 30,1998, safety evaluation. Neutron Fluence /Emb iiiiement: As published in the August 1992 Federal Register under supplementary information regarding the final rule, the NRC position with regard to augmented examination of reactor vessel shell welds is based on an embrittlement concern stemming from irradiation material test results which show that certain reactor vessel materials undergo greater radiation damage than previously expected. The BWR Vessel and Internals Project report (BWRVIP-05), dated September 1995, stated that " Embrittlement issues are addressed in 10CFR50 Appendix G through requirements associated with upper shelf energy (USE) and the reference temperature of nil-ductility transition (RT.7). In order to account for the effects of embrittlement, adjusted reference temperatures (ARTS), defined as the initial RTm plus the irradiation shift for fluence, are determined. It is possible that ARTS may result in pressure-temperature testing criteria that are difficult to meet due to increased temperature requirements. However, due to low . BWR fluence, an unacceptable ART will not be reached, even when extended life is planned." Also, the report states that "In addition to increasing RT., the USE of low alloy steel materials decreases with neutron exposure. However, for the relatively low fluence BWR, maintaining a USE above 50 ft-lbs is not a concern. Also, Code margins required by appendix G are satisfied at USE values as low as 35 ft-lbs and thus is not a safety concern. Based on the above, it can be seen that although irradiation embrittlement of materials can be a significant concern, its effect is minimal for the relatively low fluence environment of a B W R." Probabilistic Fracture Mechanics (PFM) Analvsis: Although BWRVIP-05 provides a technical basis for this relief, an independent NRC risk informed assessment of the analysis contained in the BWRVIP-05 report was conducted. The independent NRC assessment used the FAVOR code to perform probabilistic fracture mechanics (PFM) analy6is to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are: the neutron fluence was estimated to be end-of-license mean fluence, the chemistry values are mean values based on vessel types, and the potential for beyond design basis events was considered. The following is a statement contained in the " Executive Summary" of the "NRC Staff Final Safety Evaluation of BWRVIP-5 Report (Reference 3). "It should be noted that the failure frequency for axial welds cited above are relatively high, but that there are known conservatisms in these estimates. For example, these analyses were based on the assumption that the flaws in axial weld with the limiting material properties and chemistry are all located at the inside surface of the BWR RPV and at the location of peak end-of-license (EOL) azimuth fluence. Since flaws are distributed throughout the weld and EOL

Attichm:mt 1 JPN-99-025 Page 3 of 7 i n,eutron fluence will not occur for many years, the staff has concluded that the present RPV failure frequency is substantially below that reported by the BWRVIP, and independently calculated by the staff, and is not a near-term safety concern." The following information is provided to show the conservatism of the NRC analysis with respect to the FitzPatrick plant. Changes in RTuor may be used as one of the means for monitoring radiation embrittlement of reactor vessel materials. For plants with RPVs fabricated by Combustion Engineering (CE), the mean end-of-license neutron fluence for circumferential welds used in the NRC staff and BWRVIP Limiting Plant-Specific Analysis - (32 EFPY), Table 2.6-4 of the Safety Evaluation for BWRVIP-05, was 0.20E+ 19 n/cm'. However the highest fluence anticipated for FitzPatrick's belt-line circumferential weld at 2 the end of license (32 EFPY) is.161E + 19 n/cm (Reference 6). The projected fluence for the FitzPatrick plant.for 32 EFPY is less than that used in the NRC analysis. Therefore, there is significant conservatism, with regard to the effect of fluence on embrittlement, in the already low circumferential weld failure probabilities as related to the FitzPatrick NPP. Table 1 shows a comparison between the NRC Final Evaluation of the BWRVIP-05 Limiting Plant Specific Analysis data and JAF specific data for weld chemistry factor (CF), adjustment of the reference temperature (ARTuo7), and mean RTuo7 TABLE 1 i Parameter Description FitzPatrick Comparative USNRC Limiting Plant Specific Data at 32 EFPY Analysis Data at 32 EFPY, (Bounding Circ Weld) SE Table 2.6-4 CE ~BWRVIP" CE "CEOG" 2 Fluenee, n/cm 1.61 E + 18 (ID) 2.OE + 18 2.OE + 18 Initial RTuor 'F -50 (Ref. 6) O O Chemistry Factor 209.1(Ref. 6) 151.7 172.2 Cu % .337 (Ref. 6) 0.13 0.183 Ni % .609 (Ref. 6) 0.71 0.704 ARTuov'F 93.9 (Ref. 6) 86.4 98.1 Mean RTuov'F 43.9 86.4 98.1

• Per Regulatory Guide 1.99, Revision 2 Comparison of the FitzPatrick specific data and the data used in the NRC Final Safety i

Evaluation indicates differences in the combined effects of the Cu% and the Ni% on the l chemistry factor and a difference in initial RTuor. As shown above, the impact of irradiation l results in lower plant-specific mean RTwor for the FitzPatrick NPP circumferential weld material as compared to that for any of the Staff's plant-specific analyses that were performed for the CE fabricated RPV's with the highest adjusted reference temperatures. Therefore, based on plant specific data, there is a lower conditional probability of failure for circumferential welds at FitzPatrick than that stated in the NRC's Final Safety Evaluation of the BWRVIP-05. r

Attachmsnt i JPN-99-025 Page 4 of 7 2. The following response provides the basis that demonstrates that the James A. FitzPatrick has implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staff's July 30,1998, safety evaluation. At an industry 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. Later, in a Request for Additional Information (RAI) to the BWRVIP, the NRC requested that the BWRVIP evaluate the potential for a non-design basis cold over-pressure transients (Reference 4) and responded to in BWRVIP letter to NRC dated J December 18,1997 (Reference 5). The NRC also considered beyond design basis events, such as low temperature over pressure (LTOP) events in their PFM analysis, in the BWRVIP response to the RAI the total probability of an occurrence of cold overpressure for BWR-4s was reported as 9E-4. It was concluded that it is highly unlikely that a BWR would experience a cold over-pressure transient. In fact, for a BWR to experience such an - event would generally require several operator errors. 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. BWR. Also, 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. The following addresses the high-pressure injection sources, administrative controls, and operator training regarding a cold overpressure event for the FitzPatrick plant. Review of Potential High Pressure injection 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 over-pressurization event while the unit is in cold shutdown. During reactor cold shutdown conditions the condensate booster pumps are normally maintained in the " pull-to-lock" position and the feedpump 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 condensate booster pump and inject into the vessel. As discussed below, operating procedural restrictions, operator training, and work control processes at JAFNPP 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

t 3 lt - Attachmtnt 1 JPN-99-025 i Page 5 of 7 l l tp react,to unanticipated level changes and, thus, significantly reduces the possibility of an i event that would result in a violation of the pressure-temperature limits. L l 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 I would give the Operator ample time to control reactor pressure in the case of an j inadvertent injection. i 1 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 level of safety within acceptable limits. The pressure test l is conducted so that the required temperature bands for the pressure increases are achieved and maintained prior to increasing pressure. During performance of an RPV l pressure test, level and pressure are controlled using the CRD and RWCU systems using a q i " 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. l Procedural Controls / Operator Training to Prevent Reactor Pressure Vessel Cold Over-Pressurization: Operating procedural restrictions, operator training, and work control processes at JAFNPP 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 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. 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 temperature from a specified band will be promptly identified and corrected. Finally, 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 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, 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. l

y AttachmGnt i JPN-99-025 Page 6 of 7 During plant outages the work control processes ensure that the outage schedule and { changes to the schedule receive a thorough shutdown risk assessment review to ensure defense-in-depth is maintained. At JAFNPP outage work requests 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 temperature. From the outage schedule, a daily schedule is developed listing the work activities to be performed. These daily schedules - are reviewed and approved 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 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 briefings are attended by 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. Based upon the above, the probability of a low temperature over-pressure event at FitzPatrick is considered to be less than or equal to that used in the USNRC safety evaluation. Conclusion Deferral of the RPV circumferential shell weld exams to the end of the current operating license will ensure a high degree of quality and safety. Based on the documentation in the BWRVIP-05 report, the risk-informed independent assessment performed by the NRC staff, the lower neutron fluence, the less challenging design and operational loading for BWRs, the quality of the original vessel fabrication, the lack of significant degradation mechanisms, and controls to prevent a cold over-pressure event, the Authority believes a permanent deferral in performing the augmented inspections of the RPV circumferential shell welds for the remaining operating license of the plant provides an acceptable level of quality and safety. E. ALTERNATIVE EXAMINATIONS: The augmented circumferential weld inspections would be permanently deferred for the remainder of the current operating license of the plant. The alternative plan would require performance of RPV vertical weld examinations and incidental examination of 2 to 3 percent of the intersecting circumferential shell welds to the maximum extent possible ' based on accessibility. The deferral is an alternative plan to the augmented examination requirements for RPV shell welds specified in 10 CFR 50.55a(g)(6)(ii)(A)(2).

Att:chmtnt 1 JPN-99-025 Page 7 of 7 Referen.ces:

1. NRC Generic Letter 98-05, " Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," dated November 10,1998.

2. EPRI TR-105697, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), September 1995.

3. NRC Letter from Gus C. Lainas, Acting Director, Division of Engineering, Office of Nuclear regulatory Regulation, to Carl Terry, BWRVIP Chairman, Niagara Mohawk l

Company, " Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, dated July 28,1998.

4. NRC Letter from Brian W. Sheron, Director, Division of Engineering, Office of Nuclear regulatory Regulation, to Carl Terry, BWRVIP Chairman, Niagara Mohawk Company,

" transmittal of NRC Staff's Independent Assessment of the Boiling Water Reactor l Vessel and Internals Project BWRVIP-05 Report and Proprietary Request for Additional l Information", dated August 14,1997. l

5. BWRVIP Letter, Carl Terry, BWRVIP Chairman, Niagara Mohawk Company, to the NRC, C. E. Carpenter, "BWRVIP Response to NRC Request for Additional Information on BWRVIP-05", dated December 18,1997.

l

6. JAF-ICD-RPV-03393, BWRVIP Integrated Surveillance Program Data Verification Activity, Revision 0, dated July 26,1999.

7. NYPA Letter (JPN-98-039), " Request for Additional Information Regarding Response to Generic Letter 92-01: Reactor Pressure Vessel Integrity (TAC No. MA1190), dated August 31,1998.

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