ML20137R607
| ML20137R607 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/09/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137R599 | List: |
| References | |
| NUDOCS 9704140185 | |
| Download: ML20137R607 (9) | |
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UNITED STATES.
[
S NUCLEAR REGULATORY COMMISSION In f
WASHINGTON, D.C. 20665-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 222 TO FACILITY OPERATING LICENSE NO. DP 4
AND AMENDMENT N0. 213 TO FACILITY OPERATING LICENSE NO. DPR-79 s
IENNESSEE VALLEY AUTHORITY SE000YAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
The Tennessee Valley Authority (TVA, the licensee) requested amendments to Operating Licenses DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SQN), Units 1 and 2, respectively, in a letter dated October 18,.1996 as supplemented on
' March 12, March 17, April 4, and April 9, 1997.
The amendments would revise Technical Specifications (TS) 3.4.6.2, 4.4.5.4, and 4.4.5.5 and their associated Bases by permanently incorporating the requirements associated with steam generator tube inspections and repair at SQN.
These requirements establish alternate steam generator tube plugging criteria (APC) at~the tube support plate intersections.
These revised criteria had been incorporated into the TS by. previous amendments to each the operating licenses but only for Operating Cycle 8.
The proposed amendments would remove the reference to Cycle 8, thereby making the requirements applicable to all future operating cycles.
No new or different requirements are imposed by these amendments.
The alternate repair criteria, previously approved for only one operating cycle,. allow steam generator tubes having outside diameter stress corrosion 4
cracking (ODSCC) that is predominately axially oriented and confined within the tube support voltage response. plates to remain in service on the basis of bobbin coil The U. S. Nuclear Regulatory Commission (NRC) guidance on the alternate repair criteria is specified in Generic Letter (GL) 95-05,
" Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."
j The March 12, March 17, April 4, and April 9, 1997, letters provided clarifying informaton that did not change the scope of the October 18, 1996, application and the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
l The acceptance criteria (i.e., plugging limits) for steam generator tubes are specified in the plant TS.
The traditional strategy for achieving adequate structural and leakage integrity of the tubes has been to establish a minimum wall thickness requirement in accordance with NRC Regulatory Guide (RG) 1.121,
" Bases for Plugging Degraded PWR { pressurized water reactor] Steam Generator Tubes."
The minimum wall thickness requirement was developed with the I
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v l assumption of a uniform thinning of the tube wall.
This assumed degradation 1
-mechanism is inherently conservative for certain forms of tube degradation.
Conservative repair limits may lead to removing degraded tubes from service that may otherwise have adequate structural and leakage integrity for further service.
i To reduce unnecessary conservatism in the minimum wall thickness requirement for certain degradation, the industry proposed voltage-based repair criteria for ODSCC confined within the thickness of.the tube support plates. The staff published several conclusions regarding voltage-based repair criteria in draft NUREG-1477, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" and in a draft GL titled " Voltage-Based Repair Criteria for Westinghouse Steam' Generator Tubes." The latter document was published for public comment in the Federal Register on August 12, 1994 (59 FR 41520). On August 3, 1995, the staff issued GL 95-05 that took into consideration public comments on the draft GL cited above, domestic operating experience under the voltage-based repair criteria, and additional data made available from i
. European nuclear power plants.
1 The guidance of GL 95-05 does not set depth-based limits on predominantly j
axially oriented ODSCC at tube support plate locations; rather, it relies on empirically derived correlations between a nondestructive inspection parameter, the bobbin coil voltage, and tube burst pressure and leak. rate.
The staff recognizes that althcugh the total tube integrity margins may be 1
reduced following application of a voltage-based repair criteria, the guidance i
in GL 95-05 ensures that structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR Part 50 and the guideline values in 10 CFR Part 100.
Since the voltage-based repair criteria do require minimum tube wall thickness, there is the possibility for tubes with through-wall cracks to remain in service.
j Because of the increased likelihood of such flaws, the staff included provisions for augmented steam generator tube inspections and restrictive operational leakage limits.
GL 95-05 specifies, in part, that: (1) the repair criteria is only applicable 1
to predominantly axially oriented ODSCC located within the bounds of the tube support plates; (2) licensees perform an evaluation to confirm that the steam generator tubes will retain adequate structural and leakage integrity from cycle to cycle; (3) licensees adhere to specific inspection criteria to ensure consistency in methods between inspections; (4) tubes must be periodically removed from the steam generators, examined, and destructively tested to verify the morphology of the degradation and provide additional data for structural and leakage integrity evaluations; (5) the operational leakage limit be reduced; (6) licensees implement an operational leakage monitoring j
program; and (7) specific reporting requirements be incorporated into the plant technical specifications.
The staff approved the licensee's interim repair criteria (or APC) for SQN Unit I as documented in license amendment No. 214, issued on October 11, 1995, and for Unit 2 as documented in license amendment No. 211, issued on April 3, 1996.
The APC were considered interim because the amendments was approved for h
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, only one operating cycle while resolution of a number of technical issues was in the resolution process with the Nuclear Energy Institute. The APC are made permanent with these amendments and will eliminate the need for periodic j
license amendment applications in the future.
Each Sequoyah unit has four Westinghouse Model 51 steam generators, which use mill-annealed alloy 600 tubing. These steam generators use drilled-hole tube support plates and do not have flow distribution baffle plates.
The outside diameter and wall thickness of each tube is 7/8 inch and 0.050 inch, respectively.
i 3.0 EVALUATION The licensee stated that it will comply with the guidance in GL 95-05 for its APC.
In addition, the iicensee proposed to incorporate verbatim the model technical specifications in GL 95-05 into the SQN TS. The major issues related to the licensee's permanent implementation of _the APC are discussed
.below.
3.1 Tube Repair Limits The APC approved in these amendments (1) permit tubes having indications confined to within the thickness of the tube support plates with bobbin voltages less than or equal to 2.0 volts to remain in service; (2) permit tubes having indications confined to within the thickness of the tube support plates with bobbin voltages greater than 2.0 volts but less than or equal to the upper voltage limit to remain in service if a motorized rotating pancake coil probe or acceptable alternative inspection does not detect degradation; and (3) require tubes having indications confined to within the thickness of the tube support plates with bobbin voltages greater than the upper voltage limit be plugget or repaired.
The proposed lower voltage limit of 2.0 volts is based on the use of a correlation between the burst pressure and the bobbin coil voltage of pulled tube and model boiler data and is consistent with the recommended value l
specified in GL 95-05 for 7/8 inch steam generator tubing.
The upper voltage i
i limit is based on the lower 95 percent prediction interval of the burst pressure versus bobbin voltage correlation, adjusted for lower bound material properties evaluated at the 95 percent confidence level.
The upper voltage limit is further reduced to account for uncertainty in the nondestructive examination technique and flaw growth over the next operating cycle.
The industry periodically updates the database for burst pressure and bobbin t
voltage when the destructive test data from pulled tubes are available; therefore, the upper voltage limit may vary as additional data are incorporated into the correlation.
3.2 Inspection Issues Section~3.c.3 of Attachment I to GL 95-05 specifies guidance for probe wear.
1 The licensee proposed to use an alternative to section 3.c.3.
The alternative approach, developed throuch the Nuclear Energy Institute, specifies that if the probe does not satisfy ne voltage variability criterion for wear of i 15 j
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. percent limit before its replacement, all tubes which exhibited flaw signals with voltage responses measured at 75 percent or greater of the lower repair limit must be reinspected with a bobbin probe satisfying the i 15 percent wear standard criterion. The voltages from the reinspection should be used as the basis for tube repair.
The NRC staff completed a review of the Nuclear Energy Institute proposed alternative method and concluded that the approach is i
acceptable as discussed in a letter from Brian Sheron of the NRC to Alex Marion of the Nuclear Energy Institute dated March 18,-1996. The licensee's proposal to follow the industry approach to address probe wear is
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acceptable.
In the laboratory and field studies supporting the alternative probe wear criteria, the correlation of voltages measured by worn probes and new probes shows that for all significant voltage levels, the worn probe voltages are
.never less than 75 percent of the new probe voltage as discussed in the letter from Alex Marion of the Nuclear Energy Institute to Brian Sheron of the NRC dated January 23, 1996. However, in the 90-Day inspection report for Byron Unit I dated September 9, 1996, CommonwealtS Edison, the licensee for Byron, compared the worn probe voltage to the new probe voltage and found that the worn probe voltage was substantially less than 75 percent of the new probe voltage for a few indications. Commonwealth Edison evaluated these indications and concluded that the criteria to retest tubes with worn probe voltages above 75 percent of the repair limit is adequate and generally conservative due to the trend for worn probe voltages to exceed new probe voltages.
Comparison of the actual and projected end-of-cycle voltages did not show anything unusual attributable to the alternate probe wear criteria.
The staff concludes that the aforementioned probe wear results do not indicate an immediate need to modify the probe wear criteria developed by the industry.
However, the staff will continue to monitor probe wear in the 90-Day inspection reports.
Section 3.b.3 of Attachment I to GL 95-05 specifies that all tube support plate intersections with dent signals greater than 5 volts should be inspected with a rotating pancake coil (RPC).
Any tube with indications found at such intersections by RPC will be repaired.
If indications are circumferentially oriented or caused by primary water stress corrosion cracking (PWSCC), it may be necessary to expand the RPC sampling plan to include tubes with dents showing signals less than 5.0 volts.
Ste s generators 3 and 4 in SQN Unit I have experienced many dent signals that are greater than 5 volts such that inspecting all affected dents at the tube support plate intersections, as specified under section 3.b.3, would not be practical.
The licensee requested an exception to section 3.b.3.
The licensee proposed to initially sample 20 percent of the total dented intersections in the hot leg side of steam generators 3 and 4 with an RPC.
The sample will begin at the lowest tube support plate elevations, which has the highest probability of dented intersections, and continue to higher elevations.
If the RPC identified circumferential ODSCC or PWSCC at the dented intersections that the bobbin probe had missed, the sample will be expanded in accordance with a prescribed expansion plan.
For steam generators
- 1. and 2 in SQN Unit 1, the initial sample will be 100 percent of the total dented tube support plate intersections in the hot leg side.
I l For Unit I steam generators, the licensee stated that, using RPC, it will inspect all dent signals less than 5 volts at all tube support plate elevations (and lower tube support plates) where, based on past inspections, degradation has occurred (defining a critical area) and perform a 20 percent sample of the buffer zone to bound the affected area.
The licensee defined the buffer zone as the next higher tube support plate elevation where no degradation has been observed.
The buffer zone is created to ensure that the critical area is bounded.
The degradation (circumferential ODSCC or PWSCC not detected by bobbin coil) identified from the past inspection of the dented intersections would determine the initial sample.
Each initial sample will be determined independently.
If degradation was not identified in the past l
inspection, then a minimum 20 percent sample of the dents at the first tube support plate intersection will be examined.
During future outages, a different 20 percent sample would be inspected, such that over five outages 100 percent of the dents at this elevation would be inspected.
If indications i
are identified in the buffer zone, the sample will be expanded in accordance with the prescribed plan.
The classification in TS section 4.4.5.2 will be used to classify the inspection results in the buffer zone, excapt that, when a-sample size is less than 200, only category C-2 results apply The licensee also proposed an alternative to the aforementioned inspection plan for dent signals greater than 5 volts for steam generators 3 and 4 in Unit 1.
The alternative plan would follow the proposed inspection plan for dents less than 5 volts.
In addition, if a tube support plate elevation has less than 250 dented intersections when selecting a buffer zone, additional dented intersections at the next higher elevation will be inspected to make the total number of dented intersections to be inspected equal to 50.
Certain commitments were made for SQN Unit 1 in the licensee's letter dated March 12, 1997.
The licensee stated that the inspection frequency for dented i
intersections will be performed coinciding with steam generator surveillance requirements.
If an unscheduled mid-cycle steam generator surveillance is required, the dented intersections will be inspected.
The licensee will inspect the dented intersections with a technique qualified to Appendix H of the Steam Generator Examination Guideline; published by the Electric Power 4
Research Institute.
The licensee stated that any indications identified that l
exceed the plugging limit will be repaired.
Until a technique is qualified for sizing and validated for site specific applicability, tubes with PWSCC or ODSCC circumferaatial indications at dented intersections will be plugged on detection.
The licensee will follow the guidelines in GL 95-05 to inspect all dent signals in Unit 2 steam generators.
The proposed inspection plan for dent signals greater than 5 volts in Unit I steam generators is the same as the one that the staff approved in licensee amendment number 214.
For dent signals less than 5 volts in Unit I steam generators, the staff finds the proposed inspection plan acceptable because it is conservative.
The staff has no objection to the proposed alternative inspection plan for dent signals greater than 5 volts because the sample would result in an adequate number of dent intersections being inspected to identify defected tubes and to expand the-inspection as appropriate.
The inspection i
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..e plan for all dented intersections in Unit 2 steam generators follows the GL and, therefore, is acceptable.
3.3 Structural and Leakaae Intearity Assessments The staff guidance for the implementation of the voltage-based repair criteria focuses on maintaining tube structural integrity during the full range of normal, transient and postulated accident conditions with adequate allowance for eddy current test uncertainty and flaw growth projected to occur during the next operating cycle. Tube structural limits based on RG 1.121 criteria require maintaining a margin of safety of 1.43 against tube failure under postulated accident conditions and maintaining a margin of safety of 3 against burst during normal operation.
Because GL 95-05 addresses tubes affected with ODSCC confined to within the thickness of the tube support plate during normal
' operation, the staff concluded that the structural constraint provided by the tube support plate ensures all tubes to which the voltage-based criteria applies will retain a margin of 3 with respect to burst under normal operating conditions.
For a postulated main steam line break accident, however, the tube support plate may displace axially durirg steam generator blowdown such that the ODSCC affected portion of the tubing may no longer be fully constrained by the tube support plate. Accordingly, it is appropriate to consider the ODSCC affected regions of the tubes as free standing tubes for the purpose of assessing burst integrity under postulated main steam line break conditions.
In order to confirm the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies a methodology to determine the conditional burst probability and the total primary-to-secondary leak rate from an affected steam generator during a postulated main steam line break event.
To complete GL 95-05 prescribed assessments, the licensee proposes to follow the methodology described in WCAP-14277 Revision 1, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," dated December 1996.
The staff finds the methodology in WCAP-14277, Revision 1, acceptable.
GL 95-05 specifies that the structural and leakage integrity assessments should use the latest available data from destructive examinations of tubes removed from Westinghouse-designed steam generators.
The licensee stated that it will use NRC approved database.
For the upcoming cycle 8 inspection at Unit 1, the licensee will use the database previously approved by the staff in the licensee amendment No. 211 for Unit 2 dated April 3, 1996, including data from any additional pull tubes in accordance with exclusion criteria protocol in GL 95-05.
For the long term, the staff is reviewing the industry data which is documented in the EPRI report, " Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates--Database for Alternate Repair Limits, 1996 Database Update, Addendum 1," NP-7480-L, Hovember 1996.
The staff is also working with the Nuclear Energy Institute to finalize the protocol for the periodical update of the database.
The licensee stated that Sequoyah intends to use the database and to follow the protocols for future outages once they are approved.
The staff finds that the licensee's intentions to use NRC-approved database to perform structural and leakage assessments will enhance its steam generator inspection program.
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3.3.1 (onditional Probability of Burst 4
The licensee will use the methodology described in Revision 1 of WCAP-14277 for performing a probabilistic analysis to quantify the potential for steam
. generator tube ruptures given an main steam line break event.
The result the probabilistic analysis will be compared to a threshold value of 1x10',s of per cycle in accordance with GL 95-05. This threshold value p.'ovides assurance that'the probability of burst is acceptable considering the assumptions of.the calculation and the results of the staff's generic risk assessment for steam generators contained in NUREG-0844, "NRC Integrated Program for the Resolution 1
of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure to meet the threshold value indicates ODSCC confined to within the thickness of the tube support plate could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation assumed and evaluated as acceptable in NUREG-0844.
The NRC.
staff concludes the licensee's proposed methodology for calculating the conditional burst probability is consistent with the guidance in GL 95-05 and is acceptable.
3.3.2 Accident Leakaae The licensee will use the methodology described in Revision 1 of WCAP-14277 for calculating the steam generator tube leakage from the faulted steam generator during a postulated main steam line break event.
The model consists of two major components:
(1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of leakage model); and (2) a model predicting leak rate as a function of voltage, given that leakage occurs (i.e., the conditional leak rate model).
The staff concludes that the licensee's proposed methodology for calculating the tube leakage is consistent with the guidance in GL 95-05 and is acceptable.
3.3.3 Primary-to-Secondary Leakaae Durina Normal Operation Because the voltage-based repair criteria would allow degraded tubes to remain in service, the degraded tubes may develop through-wall cracks during an operational cycle, thus creating the potential for primary-to-secondary leakage during normal operation, transients, or postulated accidents.
Therefore, as a defense-in-depth measure, GL 95-05 specifies that the operational leakage limits of-the plant TS be limited to 150 gallons per day from any one steam generator.
The staff concludes that adequate leakage integrity during normal operation is reasonably assured by the TS limits on allowable primary-to-secondary leakage.
Sequoyah Units 1 and 2 TS limit the j
primary-to-secondary leakage through one steam generator to 150 gallons per day.
The staff finds that the leakage require. ment in the Sequoyah TS is consistent with the guidance in GL 95-05 and is, therefore,-acceptable.
J 3.4 Dearadation Monitorina To confirm the nature of the degradation at the tube support plate elevations, tubes are periodically removed from the steam generators for destructive tests.
The test data from removed tubes can confirm that the nature of the degradation observed at these locations is predominantly axially oriented ODSCC, provide data for assessing the reliability of the inspection methods, and supplement the existing databases (e.g., burst pressure, probability of leakage, and leak rate).
GL 95-05 specifies that at least two tubes be removed from steam generators with the objective of retrieving as many intersections as practical (minimum of four intersections) during the plant steam generator inspection outage preceding initial application of the voltage-based repair criteria. On an ongoing bases, additional tube. specimens (minimum of two intersections) should be removed at the first refueling outage following 34 effective full power months of operation or at the maximum interval of three refueling outages after the previous tube pull.
Alternatively, the licensee may participate in an industry-sponsored tube pull program endorsed by the staff as described in GL 95-05.
The licensee removed two tubes during cycle 6 outage and three tubes during cycle 7 outage from the Unit I steam generators and two tubes during cycle 7 outage from the Unit 2 steam generators for burst and leak rate testing and metallographic examination. The metallurgical examination confirmed that the degradation mechanism for the indications at the tube support plates was predominantly axially oriented ODSCC.
The licensee stated that it will comply with GL 95-05 for future tube removal. The staff concludes that the licensee satisfies the tube removal guidance of GL 95-05.
4.0
SUMMARY
The licensee submitted an application for a license amendment to permit the use of the voltage-based repair criteria for steam generator tubes at Sequoyah Units 1 and 2 on a permanent basis.
This would be accomplished by removing the footnote on each affected TS page that resticted the changes to Cycle 8 only (e.g., "The indicated changes to this page are applicable to Cycle 8 operation only").
The staff has reviewed the proposed amendment and concludes that the proposed alternate repair criteria are consistent with GL 95-05 and, therefore, the proposed TS changes are acceptable.
The staff also concludes that adequate structural and leakage integrity can be assured, consistent with applicable regulatory requirements, for indications to which the voltage-based repair criteria will be applied.
The staff approves the proposed voltage-based repair criteria based in part on the licensee being able to successfully demonstrate after each inspection outage that the conditional probability of burst and the primary-to-secondary leakage during a postulated main steam line break will be acceptable per the guidance in GL 95-05.
The staff's approval of this amendment request relies upon information provided by TVA in their letters related to this request dated October 18, 1996, March 12 and 17, 1997, and their concurrence with a proposed license condition on April 4, 1997.
Specifically, the commitments made by TVA in their March 12, 1997, letter have been incorporated into the " Steam Generator Inspection" license condition section (Condition 2.C.9 for Unit I and 2.C.8 for Unit 2) as discussed in the license amendment.
This change is acceptable.
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5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Tennessee State official was -notified of.the proposed issuance of the: amendments.
The State official l
had no comments'.
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6.0 ENVIRONMENTAL CONSIDERATION
-The amendment changes a requirement with respect to installation or use of a
-facility component located within the restricted area'as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that.the amendment involves no significant' increase in the amounts, and no significant ' change in the types,. of any effluents that may be released offsite,,and that1there is no significant increase in individual or cumulative
- occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards l consideration, and there has been no public comment on such finding (62 FR 6276 dated February 11,1997). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 :FR i
51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or l
environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (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 defutse and security cr to the health and safety of the public.
Principal Contributor:
John C. Tsao Dated: April 9, 1997 r
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