ML20205N652
| ML20205N652 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/08/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20205N642 | List: |
| References | |
| NUDOCS 9904190060 | |
| Download: ML20205N652 (5) | |
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UNITED STATES j
I2f. j NUCLEAR REGULATORY COMMISSION
"' f WASHINGTON, D.C. 20555 4 001
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.172TO FACILITY OPERATING LICENSE NO. DPR 72
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FLORIDA POWER CORPORATION CRYSTAL RIVER.. UNIT 3 DOCKET NO. 50-302
1.0 INTRODUCTION
By letter dated October 1,1997, as supplemented on April 23 and November 17,1998, and February 19,1999, Florida Power Corporation (the licensee) submitted proposed changes to i
the improved Technical Specifications (ITSs) for Crystal River Unit 3 (CR-3). The changes specify citeria for evaluating the growth of pit-like intergranular attack (IGA) steam gcnerator tube degradation identified in tubes in the "B" once-through steam generator (OTSG). The licensee has also requested to amend the ITSs to clarify the date by which the OTSG inservice inspection results are required to be submitted to the U.S. Nuclear Regulatory Comrnission (NRC). The April 23 and November 17,1998, and February 19,1999 letters did not affect the original no significant hazards determination. The following documents the staff's assessment i
of the changes proposed by the licensee.
2.0 BACKGROUND
2.1 Reaulatory Framework For Proposed Licensino Action Steam generator tubing comprises a significant fraction of the reactor coolant pressure boundary. Title 10 of the Code of Federal Regulations (CFR),10 CFR 50.55a(c) specifies that components that are part of the reactor coolant pressure boundary must be designed and i
constructed to meet the requirements for Class 1 components in Section ill of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. To ensure the continued integrity of the tubing at operating pressurized water reactor (PWR) facilities,50.55a further rcquires that throughout the service life of a PWR facility, Class 1 components meet the requirements in Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components" of the ASME Code. This requirement includes the inspe'. ion and tube repair criteria of Section XI of the ASME Code. However, an exception is provided for design and access provisions and preservice examination requirements in Sction XI. In addition,10 CFR 50.55a(b)(2)(iii) states that if the technical specification (TS) surveillance requirements for steam generators differ from those in Article IWB-2000 of Section XI of the ASME Code, the inservice inspection program is governed by the TSs.
As part of the plant licensing basis, applicants for a PWR operating license analyze the consequences of postulated design basis accidents that assume degradation of the steam 9904190060 990400 PDR ADOCK 05000302 P
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- generator tubes such that primary coolant leaks to the secondary coolant side of the steam generators. Examples of such accidents are a steam generator tube rupture, s main steam line break, a locked rotor, and a control rod ejection. Analyses of these accidents consider the
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primary-to-secondary leakage that may occur during these postulated events when demonstrating that radiological consequences do not exceed the 10 CFR Part 100 guidelines, or some fraction thereof, for offsite doses, nor General Design Criterion 19 for control room operator dosos. The staff uses criteria specified in NUREG-0800, the Standard Review Plan, to evaluato these accidents.
A plant's TSs require that licensees perform periodic inservice inspections of the steam generator tubing and repair or remove from service (by installing plugs in the tube ends) all tubes exceeding the tube repair limit. In addition, operational leakage limits are included in the TSs to ensure that, should tube leakage develop, the licensee will take prompt action to avoid rupture of the leaking tubes. These requirements are intended to ensure that burst margins are maintained consistent with Appendices A and B to 10 CFR Part 50 and that the potential for leakage is maintained consistent with what has been analyzed as part of the plant licensing basis.
Revision 1 of NRC Regulatory Guide (RG) 1.83," Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," provides guidance conceming steam generator tube inspectior scope and frequency and nondestructive examination methodology. RG 1.83 is referenced,a the standard review plan and is intended to provide a basis for reviewing inservice inspection criteria in the TSs. NRC Regulatory Guide 1,121," Bases for Plugging Degraded PWR Steam Generator Tubes," provides guidelines for determining the tube repair criteria and operationalleakage limits that are specified in the TSs. Together, these two RGs provide specific criteria that should be considered in proposed steam generator tube alternate repair criteria in order to satisfy the previously mentioned regulatory requirements.
2.2 Description nf Proposed Amendment A number of tubes have been identified as containing apparent indications of pit-like IGA degradation in the "B" OTSG at CR-3 for a number of operating cycles. The licensee has attempted to develop attemate repair criteria to address this mode of degradation in previous cycles. However, the NRC has not accepted previous proposals developed by t% licensee to disposition tubes with pit-like IGA indications on a permanent basis. In the 1997 inspections at CR-3, the licensee applied a method to estimate the depth of IGA degradation using traditional eddy current inspection techniques. The technique relied on using inspection data obtained with high frequency bobbin coil probes. Tubes with IGA flaws witn depths measured in excess of the existing depth-based tube plugging limits in ITS 5.6.2.10.4.a.7 (i.e.,40 percent of nominal tube wall thickness) were removed from service. In order to ensure adequate steam generator tube integrity through the end of the operating cycle following the outage in which tubes are inspected, licensees typically complete operational assessments to conservatively estimate the conditional probability of tube rupture and steam line break primary-to-secondary leak rate through d*ctive steam generator tubes. One critical factor in these assessments is the accuracy of growth rates for known degradation mechanisms.
The licensee has been inspecting tubes with pit-like IGA indications for a number of cycles.
Assessments of IGA flaw growth rate completed by the licensee have concluded that this
3-damage mechanism is dormant for flaws in the lower bundle region of tho "B" OTSG. The staff has previously concluded that changes in the eddy current data acquisition equipment use.d in successive inspections may have hindered the licensee's ability to detect minor changes in flaw growth rate. Such changes could have altered the eddy current readings for flaws inspected during steam generator tube examinations. Thereforn, the staff has not accopted the licensee's conclusion on pit-like IGA flaw progression. In addition, previous growth rate assessments were largely based on voltage change. The current approach used at CR-3 to disposition tubes with these indications is based on flaw depth. As such, overall grcwth rates will be determined on the basis of flaw depth rather than voltage in future inspections.
The licensee currently inspects the CR-3 steam generators using a mid-range frequency bobbin coil eddy cunent probe. Potential IGA indicttions detected during these inspections are reinspected using rotating probes to characterize the mode of degradation and a high frequency bobbin coil probe to estimate the through-wall depth of pit-like IGA degradation.
During the 1997 steam generator tube examinations, the licensee established a baseline by which future. degradation growth would be measured. In' order to accurately assess any changes in the depth of degradation identified and measured in the baseline examination, subsequent examinations will use data acquisition equipment that is equivalent to the system used when the depth sizing technique was qualified prior to the 1997 inspections. The proposed ITS changes would establish this requirement for evaluating growth rates.
Specifically, the growth of indications identified in future inspections will be determined by comparing the examination results to those obtained in the 1997 inspection using equivalent test probes.
The current CR-3 ITSs require the licensee to submit to the NRC the results of its OTSG inservice inspection within 90 days after completion of the inspections (ITS 5.7.2, "Special Reports"). The licensee has proposed to modify thic requirement to mandate the submittal of this report within 90 days after restarting the unit (i.e., breaker closure). The purpose for this proposed change is to more clearly define the timing when the report is required to be sent to the NRC and to specify a requirement consistent with guidance issued by the NRC for submitting reports for generator tube altemate repair criteria.
3.0 STAFF EVALUATION The current methodology for dispositioning tubes with pit-like IGA indications ir, the "B" OTSG at CR-3 relies on the determination of flaw through-wall depth using eddy current inspection techniques. In order to evaluate changes in the depth of previously identified degradation, the licensee will compare the depth determined in the most recent inspection to that determined from inspections completed in 1997. The resulting growth rate for an indication is the difference between the two measured indication depths. Indication depth measurements are obtained using eddy current probes equivalent to those used in the 1997 baseline examination. The requirement to use equivalent test probes will reduce errors in the calculated growth rate introduced by using different eddy current probes (e.g., mia-range bobbin probe) in two differentinspections. Therefore, the staff concludes th M the proposed ITS changes to TS 5.6.2.10.c, which provide that the licensee will determirt.iaw growth rate based on through-wall depth compared with the 1997 baseline examination are acceptable because data acquisition uncertainty is minimized, and the results can be directly related to the CR-3 steam generator tube plugging limit.
t The existing ITSs for CR-3 state that the licensee shall submit complete resul;s of the OTSG inservice inspection within 90 days following completion of the inspections. The licensee has j
proposed to modify this requirement in TS 5.7.2.c by stating that this report is due 90 days after unit startup from the outage (i.e., breaker closure). This change would slightly extend the date for submitting this report, but the revision would update the requirements at CR-3 to be consistent with these adopted by other PWR facilities. Any additional delay in submitting this report resulting from this change should not impact the NRC staff's ability to adequately assess any steam generator tube integrity issues stemming from the inspections in a reasonable period of time. la addition, other reporting requirements specified in the ITSs and 10 CFR 50.72 and 50.73 enable the NRC to stay informed of any identified conditions that could affect the health and safety of the public.
Therefore, the staff concludes that the proposed change to specify that a complete inservice inspection report is due 90 days from restart is acceptable.
4.0 STATE CONSULTATION
Based upon a letter dated March 8,1991, from Mary E. Clark of the State of Florida, Department of Health and Rehabilitative Services, to Deborah A. Miller, Licensing Assistant, U.S. NRC, the State of Florida does not desire notification of issuance of license amendments.
5.0 ENVIRONMENTAL CONSIDERATION
S The amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve.no significant increase in the arnounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (62 FR 54873). The amendment also changes reporting or record-keeping requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
Based on its review of the licensee's proposal, the staff has determined that the proposed changes to the CR-3 ITSs will continue to provide adequate assurance of steam generator tube integrity because the growth rate of pit-like IGA degradation will be monitored in a manner that will minimize uncertainties introduced by the data acquisition equipment. In addition, the modification to the reporting requirements will not affect the prompt notification of the NRC of any potentially significant issues related to the inservice inspection of the OTSG tubing because alternate reporting requirements will ensure prompt notification of the staff of potentially safety significant issues. The staff concludes that (1) there is reasonab8e assurance that the health and safety of the public will not be endangered by operation in the proposed mannw, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the hea!th and safety of the public.
Principal Contributor: Phillip J. Rush, DE/EMCB Date: April 8, 1999
.J-Mr., John Paul Cowan CRYSTAL RIVER UNIT NO. 3 Florida Power Corporation cc:
Mr. R. Alexander Glenn Ms. Sherry L. Bernhoft, Director Corporate Counsel Nuclear Regulatory Affairs (SA2A)
Florida Power Corporation Florida Power Corporation MAC-ASA Crystal River Energy Complex P.O. Box 14042 15760 W. Power Line Street
- St. Petersburg, Florida 33733-4042 Crystal River, Florida 34428-6708 Mr. Charles G. Pardee, Director Senior Resident Inspector Nuclear Plant Operations (NA2C)
Crystal River Unit 3 Florida Power Corporation U.S. Nuclear Regulatory Commission Crystal River Energy Complex 6745 N. Tallahassee Road 15760 W. Power Line Street Crystal River, Florida 34428 Crystal River, Florida 34428-6708 Mr. Gregory H. Halnon Mr. Michael A. Schoppman Director, Quality Programs (SA2C)
Framatome Technologies Inc.
Florida Power Corpo'ation 1700 Rockville Pike, Suite 525 Crystal River Energy Complex Rockville, Maryland 20852 15760 W. Power Line Street Crystal River, Florida 34428-6708 Mr. William A. Passetti, Chief Department of Health Bureau of Radiation Control 2020 Capital Circlel, SE, Bin #C21 Tallahassee, Florida 32399-1741 Attoiney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Chairman Board of County Commissioners Citrus County 110 North Apopka Avenue inverness, Florida 34450-4245