ML19324B800
| ML19324B800 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/30/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19324B797 | List: |
| References | |
| IEB-88-002, IEB-88-2, NUDOCS 8911080260 | |
| Download: ML19324B800 (4) | |
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UNITED STATES
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l SAFETY EVALUATION BY ThE OFFICE OF KUCLEAR REACTOR REGULATION l
RELATED TO CLOSE0VT OF BULLETIN 68 02 ISSUES (MPA B-099) t ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 (FARLEY)
DOCKET NOS. 50-348 AND 50 064
1.0 INTRODUCTION
Alebama Power Company (the licensee) submitted its response to NRC Bulletin 88-02, "Ra > idly Propagating Fatigue Cracks in Steam Generator Tubes" by letters i
dated Marcs 23 and August 29, 1988.Bulletin 88-02 requested that licensees i
for plants with Westinghouse steam generators employing carbon steel support l
plates take certain actions (specified in the Bulletin) to minimize the j
potential for 4, steam generator tube rupture event caused by a rapidly propagating fatigue crack such as occurred at North Anna Unit 1 on July 15, i
1987.
i 2.0 DISCUS $10N The licensee reports that the Farley steam generators exhibit evidence of i
denting at the uppermost support plate.
Accordingly, Items C.1 and C.2 of the bulletin are applicable to Farley.
In accordance with Item C.1 of the Bulletin, the licensee has implemented an enhanced primary-to-secondary leak rate monitoring program which is described in the licensee's March 23, 1988 letter.
This enhanced leak rate monitoring program is an interim compensatory measure sending completion of the actions requested in Item C.2 of the Bulletin and MC staff review and approval of these actions.
l The licensee het implemented the generic program developed by Westinghouse to i
resolve item C.2 of the Bulletin.
The licensee's implementation of this program is described in Westinghouse reports WCAP-11875 (Proprietary Version) and WCAP-11876 (Non-Proprietary Version) which were sut,mitted with the licensee's letter dated August 29, 1988. These reports describe the analyses which were conducted to establish the susceptibility of the Farley steam generator tubes to rapidly propagating fatigue cracks and to identify any needed corrective actions.
The staff has reviewed the Westinghouse generic program and documented its evaluation in Reference 1.
The staff concluded in Reference 1 that the Westinghouse program is an acceptable approach for resolving Item C.2 of the Bulletin.
The staff further concluded that the Kestinghouse program, if properly implemented, will prcvide reasonable assurance against further 8911080260 891030 PDR ADOCK 05000348 Q
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l f ailures of the kind which occurred at t! orth Anna Unit 1.
The Safety Evaluation herein incorporates the staff's goreric Reference 1 evaluation by reference.
The analyses for the Farley steam generators constrvatively assuned that all j
unsupported tubes are dented at the uppermost support plate.
In additicn, the l
stress ratio and fatigue estimates were based on the assumption of a full nean t
stress effect (i.e., yield stress), consistent with staff findings in Reference i
1.
i Stability ratios for the Farley steam generator tubes were determined from detailed analyses performed for another plant with bcdel 51 steam generaturs l
with very similar thernal-hydraulic conditions as Farley.
These analyses included tube instability artlyses performed with the FASTVlb computer code using thermal-hydraulic input from a 3-D ATHOS model. The stability ratio results were adjusted upward by 2.4% to reflect the slightly different i
thermal-hydraulic conditions at Farley.
The adjustnent factor was calculated t
on the b6 sis of a com arison of stability ratio estin.ates for both plcnts dettrmined by 1 D ana ysis.
The original anti-vibration bar (AVB) supports in the Farley steam generators were replaced with a modified design between 1985 and 1987.
Evaluation of the eddy current data revealed the modified AVDs to have very uniform insertion depths (i.e., depth variations are typically within one tube pitch) with the bottom of the AVBs being located between rows 11 and 12.
Because of the relatively uniform insertion depths, flow peaking factors for the Farley steem generator tubes were determined to be small compared to peaking factors which exist for certain tubes at other plants (including the tube that failed at tiorth Anna) which continue to employ the original AVE design.
The analyses docunented in WCAF-11875 show that all presently unsupported tubes in tt.e Farley steam generators satisfy the Westinghouse stress ratio criterion.
The fatigue usage factor for the mcst limiting tube is calculated to be 0.4 l
from the time cf /VB replacen.ent (1987) to the projected 40 year lifetine of l
the plant.
Fatigue usage prier to AVB replecoment was not calculated since the t
l old tVB positicos and associated peaking f actors wcre not determined.
Westinghouse notes, however, that experience at other plants indicates that ftw t
tubes in rows 10 and 11 were likely to have been not supported by AYB's prior to acdification of the AVBs.
Further, it has been Westinghouse's experience that only a small portion of unsupported tubes have exhibited significant fh peaking factors.
Thus, Westinghcuse concludes that the probability of Row 11 and smaller Farley tubes having a significant prior fatigue usage and/or e I
large crack is very small.
Even if a crack were to initiate during subsequent operation, Westinghouse concludes that the crack growth rate woulc be insignificant since even a 30" through-wall circumferential crack would not result in crack tip stresses that are above the threshold for crack l
propagation.
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Finally, the staff notes thht an N-16 monitor has been installed en all three steam lines at Farley Unit 2 ano that similar monitors were installed at Farley
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l, Unit I during the 1989 refueling outage which began on Sestember 23, 1989.
The staff considers this to be a significant enhancement of tie licensee's capability to detect and monitor rapidly changing leak rates.
(Note that the N-16 monitor is not part of the interim measures implemented by the licensee in response to Item C.1 of Bulletin 88-02). Reacings f rom the N-16 monitor are autonatictlly converted to primary-to-secer.cary leakage rate which is continuously oisplayed in the control room.
The leak rates from the N-16 nenitor are logged each shift, In addition, alarms with setpoints corresponding to three different leak rate thresholds provide acded assurance that the operators will be alerted to a rapidly increasing primary-to-secondary leak rate.
3.0 CONCLUSION
The staff concluces that the actions taken by the licensee resolve the issues ioentified in Bulletin 88-02 and are, teerefore, acceptable. Consistent with staff finding ho. 11 in Rcference 1, the licensee's staff has advised us that existing 10 CFR 50.59 administrative procedures would ensure that updated stress ratio ano fatigue usage calculations would be performed in the event of any significant changes to the steam generator operating parameters (e.g.,
steam pressure, tlow, and circulation ratio) ralative to the reference parameters asst.c;ed in the Farley analysts.
4.0 REFERENCE 1.
Safety Evaluetion Report. " Evaluation of Westinghouse Methodology to Accress Item C.2 of NRC Bulletin 88-02" which was transmitted to Westinghcuse by letter cated October 2,1987.
Principal Contributor:
E. Hurphy Dated:
October 30, 1989 A
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i DISTRIBUTION MM n
'NRC PDR Local PDR S. Varga 14-E-4 G. Lainas 14-H-3 E. Adensam 14-B-20 P. Anderson 14-B-20 i
E. Reeves 14-B-20 OGC (For inform. Only) 15-B-18 i
E. Jordan MNBB-3302 B. Grimes 9-A-2 I
ACRS (10)
P-315 Farley File I
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