ML20129A936
| ML20129A936 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/11/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20129A903 | List: |
| References | |
| NUDOCS 9610220289 | |
| Download: ML20129A936 (5) | |
Text
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-'k UNITED STATES g
g NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 20555-0001
.....,e SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FELATED TO AMENDMENT N0.115 TO FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY. INC.
JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 DOCKET N0. 50-364
1.0 INTRODUCTION
By letter dated March 29, 1996, as supplemented by letters dated June 27, August 29, August 30 and September 16, 1996, the Southern Nuclear Operating Company, Inc., et al. (the licensee), submitted a reouest for changes' to the Josepn M. Farley Nuclear Plant, Unit 2, Technical Sp5cifications (TS).
The amendment requested staff approval of permanent voltage-based alternative repair criteria for steam generator tubes in the TS.
The alternative repair criteria allow steam generator tubes, having outside diameter stress corrosion cracking (0DSCC) that is predominately axially oriented and that is confined within the tube support plates, to remain in servi & If the tube inspection and associated results satisfy the guidance in Generic Letter (GL) 95-05,
" Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," issued on August 3, 1995.
By letters dated June 27, August 29, August 30 and September 16, 1996, the licensee submitted additional information to clarify the changes to the proposed repair criteria, which did not change the scope of the March 29, 1996, application and the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
GL 95-05 outlined acceptance criteria and limitations for licensees considering implementation of voltage-based alternate repair criteria in plant technical specifications.
The traditional regulatory framework for monitoring steam generator tube integrity establishes requirements for a minimum wall thickness in accordance with Regulatory Guide 1.121.
The plant TS require that any tubes having degradation which exceeds 40 percent through wall will be repaired.
The 40 percent limit is conservative for highly localized flaws such as pits, short cracks and in particular ODSCC that occurs at the tube support plates.
The voltage-based alternate repair criteria do not set limits on the depth of ODSCC to ensure tube integrity; instead it relies on correlating the eddy current voltage amplitude from a bobbin coil probe with the specific measurement of tube burst pressure and leak rate. This approach takes no credit for the tube support plates in preventing and/or reducing the likelihood of a tube from bursting and/or leaking during postulated accident conditions.
It assumes that the degradation affecting the steam generator tubes at the tube support plate elevation is in the free span region of the tubes.
9610220289 961011 PDR ADOCK 05000364 P
l GL 95-05 specifies, in part, that: (1) the repair criteria is only applicable to axially oriented ODSCC located within the bounds of the tube support plates; (2) licensees should perform an evaluation to confirm that the steam generator tubes will retain adequate structural and leakage integrity until the next scheduled inspection; (3) licensees should adhere to specific inspection criteria to ensure consistency in methods between inspections; (4) tubes should be periodically removed from the steam generators to verify the morpholegy of the degradation and provide additional data for structural and leakage khgrity evaluations; (5) the operational leakage limit should be reduced; @) specific reporting requirements should be followed; and (7) the licensees' proposed TS should follow the moael TS in Attachment 2 of the GL.
3.0 EVALUATION j
Farley Unit 2 uses three Westinghouse model 51 steam generators. The tubes were fabricated using mill annealed alloy 600 material Each steam generator has 3,388 tubes.
The staff approved three applications for the interim alternate repair criteria in the Farley Unit 2 TS in License faendment No. 87 that was issued on April 1, 1992; License Amendment No. 94 (.a October 20, 1993; and License Amendment No. 106 on April 7, 1935.
Each of the interim criteria is approved for specific operating cycles. The misting interim criteria will expire at the end of the current operating cy ae in October 1996.
The permanent j
alternate repair criteria will replace the interim criteria and will eliminate the need for periodic license amendments for the tube repair criteria addressed by Gl. 95-05.
The licensee has committed to follow the getdance in GL 95-0:i for its proposed permanent alternate repair criteria.
In addition, the licensee has proposed to incorporate verbatim the model TS in GL 95-05 into the Farley Unit 2 TS.
Clarifications regarding the use of GL 95-05 are discussed below.
Section 1.b.1 of Attachment 1 to GL 95-05 specifies that the repair criteria do not apply to tube-to-tube support plate interrections where the tube with degradation may potentially collapse or deform as a result of the combined postulated loss-of-coolant accident and safe shutdown earthquake loadings.
Licensees should perform or reference an analysis that identifies which intersections are to be excluded.
The licensee submitted an analysis, WCAP-12871, Revision 2, as a part of its application for the interim alternate repair criteria on May 28, 1993. As a result of the licensee's analysis, no tubes need to be excluded from application of the voltage-based repair criteria. The staff agreed with the licensee's assessment as documented in license Amendment No. 94.
Section 2.b of Attachment 1 to GL 95-05 specifies criteria for an acceptable evaluation to confirm that the tubes will maintain adequate structural and leakage integrity until the next scheduled inspection.
The evaluation includes a conditional burst probability calculation and a total leak rate calculation from the affected steam generators during a postulated main steam
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1 1
' line break.
The licensee stated that it will follow the methodology described in the Westinghouse report, "SLB Leak Rate and Tube C.est Probability Analysis Methods for 00 SCC at TSP Intersections," WCAP '4277.
The report prescribes deterministic and probabilistic methods.
The staff has approved the probabilistic method in WCAP-14277 for calcul:tieg the leak rate and tube burst probability because it is consistent with the acceptance criteria outlined in GL 95-05.
The licensee uses '.he p"obabilistic methodology to evaluate structural and leakage integrity of che tubes.
The NRR staff will periodically verify the results of these calculations and assess the effectiveness of the methodologies to ensure that they are consistent with Section 2.0 of Attr.chr.ent I to GL 95-05.
Section 2.b.3(2) of Attachment I to GL 95-05 recommends that licenrees use the latest NRC-approved industry database (e.g., burst pressure, probability of leakage, and conditional leak rate database) in their tube integrity evaluations that should include calculation of tube repair limits, ccnditional burst probability, and total leakage under postulated accident conditions.
1 The licensee stated that the latest NRC-approved database, using the NRC-approved data exclusion criteria, will be applied to the tube integrity evaluations.
For the upcoming operating cycle, the licensee committed to use the database forwarded to the NRC by Duquesne Light Company letter for Beaver Valley Unit 1, dated March 27, 1996. The staff finds that the database submitted by Duquesne Light Company is acceptable for the tube integrity calculations for the upcoming operating cycle.
For the long-term, Nuclear Energy Institute (NEI) is developing a protocol for updating the steam generator degradation database.
The staff will review the adequacy of the updating process and the associated database.
Pending the implementation of an NRC-approved process for updating a generic industry database for steam generator tube degradation, the licensee will provide, as specified in GL 95-05, the database it intends to use prior to each refueling outage.
The database will include the data from tubes that have been pulled and tested up to 2 months before the outage.
The staff finds that the licensee's use of the databate satisfies Section 2.b.3(2) of Attachment I to GL 95-05.
Section 2.c of Attachment 1 to GL 95-05 specifies an alternative for licensees to calculate the primary-to-secondary leakage and probability of tube burst given a main steam line break using the projected end-of-cycle voltage distribution.
The licensee has performed the calculations on the basis of the projected end-of-cycle distributions.
In the event that the growth rate determinations cannot be completed before returning the steam generators to service, the licensee will use the actual end-of-cycle distributions as allowed in Section 2.c.
The licensee stated that even if the calculation made before returning the steam generators to. service is based on the actual measured voltage distribution, the calculation based on the projected end-of-cycle voltage distribution will be submitted to the NRC in the 90-day report following the outage.
The licensee's calculations for the primary-to-secondary leakage and tube burst probability are consistent with Section 6.5.(c) of Attachment 1 to GL 95-05 and is acceptable.
_4_
Section 3.b of Attachment 1 to GL 95-05 specifies guidance for tube inspection using the rotating pancake coil.
The licensee stated that it will use a motorized rotating coil probe, e.g.,
the + Point coil, in addition to the rotating pancake coil.
The licensee provided this clarification to ensure that the + Point coil can be used as an alternative to the rotating pancake coil.
The staff finds that the use of the + Point coil is consistent with Note 1 on page 3 of GL 95-05 and is acceptable.
Sections 3.c.2 and 3.c.3 of Attachment 1 to GL 95-05 specify guidance in regard to probe wear and variability, which the licensee will follow.
In addition, the licensee will follow the guidance in the two letters from NEI to NRC dated January 23 and February 23, 1996, and the two letters from NRC to NEI dated February 9 and March 18, 1996.
The licensee will verify that both the primary and mix frequencies of the '~cb1 will satisfy the i 10 percent variability requirement.
The staff finds the licensee's program for probe wear and variability is consistent with GL 95-05 and the staff position.
Section 3.c.5 of Attachment I to GL 95-05 specifies quantitative noise criteria (e.g., electrical noise, tube noise, calibration standard noise) for the probe.
The licensee stated that quantitative noise criteria have historically been applied and will be incorporated in the data acquisition procedures.
This allows noise levels to be addressed at the beginning of inspection.
Probes are replaced before exceeding the noise criteria.
I f, upon measurement, the probe in use fails to meet the criteria, tubes tested with that probe since the last satisfactory measurement are reinspected.
In addition, the Farley analysis procedures allow the analyst to require reinspection due to noise on a qualitative basis.
The staff finds the licensee's use of quantitative noise criteria consistent with GL 95-05.
The proposed amendment revised TS 3/4.4.6, " Steam Generators" and associated Bases section as part of implementing the voltage-based repair criteria for steam generator tubes.
Specifically, the licensee changed the following Surveillance Requirements sections in the TS:
TS 4.4.6.2.1, " Steam Generater Tube Sample Selection and Inspection;" TS 4.4.6.4.a, " Acceptance Criteria;"
and TS 4.4.6.5, " Reports." The changes incorporate the methodology of calculating the upper voltage repair limit and mid-cycle repair limits. The licensee also changed TS Bases Section 3/4.4.6, " Steam Generator," consistent with these changes as stated above. After its review, the staff concludes that the proposed TS changes satisfy the model technical specifications for the voltage-based repair criteria as specified in GL 95-05 and, therefore, are acceptable.
4.0 STAFF CONCLUSION On the basis of the information submitted, the staff concludes that the proposed permanant alternate repair criteria for the steam generator tubes in Farley Unit 2 are consistent with GL 95-05 and are acceptable.
The staff also concludes that adequate structural and leakage integrity can be ensured, consistent with applicable regulatory requirements, fcr indications to which the voltage-based repair criteria will be applied.
The licensee may incorporate the proposed permanent alternate repair criteria into the Farley Unit 2 TS.
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6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the 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 that there is no significant increase in individual or cumulative i
occupational radiation exposure.
The Comission has previously issued a l
proposed finding that the amendment involves no significant hazards i
consideration, and there has been no public coment on such finding (61 FR 25711 dated May 22,1996). Accordingly, the amendment meets the cligibility j
criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental i
assessment need be prepared in connection with the issuance of the amendment.
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7.0 CONCLUSION
The Comission 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, 1
and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
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Principal Contributor:
Date: October 11, 1996 i