ML20005E708
| ML20005E708 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/28/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20005E707 | List: |
| References | |
| NUDOCS 9001100162 | |
| Download: ML20005E708 (3) | |
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ENCLOSURE 1 i
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
CLOSE00T OF BULLETIN 88-02 ISSUES YANKEE ATOMIC POWER COMPANY YANKEE R0WE
- l DOCKET NO.50-029
1.0 INTRODUCTION
i Yankee Atomic Power Company (the licensee) submitted its response to NRC Bulletin 88-02, " Rapidly Propagating Fatigue Cracks in Steam Generator Tubes" by letters j
dated March 25, September 1, September 16, December 1, and December 7,1988.
Bulletin 88-02 requested that licensees for plants with Westinghouse steam generators employing carbon steel support plates take certain actions (specified 1
in the bulletin).to minimize the potential for a steam generator tube rupture event' caused by a rapidly propagating fatigue crack such as occurred at North Anna Unit 1 on July 15, 1987.
2.0 DISCUSSION The licensee reports that the Yankee Rowe steam generators exhibit evidence of denting at the uppermost support plate. Accordingly, items C.1 and C.2 of the bulletin are applicable to Yankee Rowe.
In its letter dated March 25, 1988, the licensee stated that it has implemented an. enhanced primary-to-secondary leak rate monitoring program in accordance with
. item C.1 of the bulletin. This enhanced leak rate monitoring program is an interim compensatory measure pending completion of the actions requested in item C.2 of the bulletin and NRC staff review and approval of these actions.
4 The licensee has implemented the generic program developed by Westinghouse to resolve item C.2 of the bulletin. The licensee's implementation of this program is described in Westinghouse report WCAP-12079 (Proprletary) which was submitted with the licensee's letter dated December 7,1988. This report describes the analyses which were conducted to establish the susceptibility of the Yankee Rowe 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 evalu-ation 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 Westinghouse program, if properly implemented, will provide reasonable assurance against further failures of the kind which occurred at North Anna Unit 1.
The safety evaluation herein incorporates the staf f's generic Reference 1 evaluation by reference.
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/ 1 Yankee Rowe is unique among PWRs in that the steam generator tubes are fabricated from 304 stainless steel rather than Inconel 600. The corrosion fatigue proper-ties of 304 stainless steel in an AVT water environment were assumed by Westing-house be similar to those for Inconel 600 based on a detailed evaluation presented in WCAP-12079. - A staff consultant, Viking Systems International, reviewed this Westinghouse evaluation and concluded in Reference 2 that Westinghouse's use of the fatigue design curve for Inconel 600 in 600 F AVT water-(Proprietary Figure 6.2 in WCAP-12079) is acceptable for performing the Yankee Rowe tube fatigue analyses.
Fluidelastic stability ratios at Yankee Rowe relative to that for the tube which ruptured at North Arna 1 were determined on the basis of a one-dimensional analysis using recent thermal-hydraulic operating parameters for the most limit-ing steam generator. Justification for use of the 1-D technique in lieu of the 3-D technique used for most other Westinghouse PWRs was obtained by extensive comparisons of 1-D predictions with 3-D predictions for a broad spectrum of operating conditions.
From these studies and to ensure the conservatism of the analysis, the 1-D stability ratio estimates were raticed up by 15%. Based on its review, the staff concurs that this should result in conservative results for Yankee Rowe.
The AVBs in the Yankee Rowe steam generators nominally extend to row 9.
Eddy current data is available for 3 of the 4 steam generators confirming that all tubes in row 11 (and beyond) are supported by AVBs.
However, there is only limited eddy current information availablo for rows 8 through 10.
None of the available information indicates any unsupported tubes in these rows; however, AVB support cannot be assured to exist for all tubes in these rows.
Furthermore, the eddy current data is not sufficient to determine the AVB configuration and associated flow peaking factor for each tube in these rows. All tubes in these rows, therefore, have been assumed to be unsupported.
Furthermore, each tube in these rows has conservatively been assigned a flow peaking factor of 1.21 times the peaking factor estimated for the tube which ruptured at North Anna. This peaking factor assumption upper bounds the worst case peaking factors found for all AVB configurations observed in the field to date at other PWRs.
The analyses for the Yankee Rowe steam generators conservatively assumed that all unsupported tubes are dented at the upperinost support plate.
In addition, the stress ratio and fatigue estimates were based on the assumption of a full mean stress effect (i.e., yield stress), consistent with the staff findings in Reference 1.
The analyses documented in WCAP-12079 show that all unsupported tubes in the Yankee Rowe steam generators satisfy the Westinghouse stress ratio criterion.
The corresponding fatigue usage factors have been determined adequate to support a 60 year plant lifetime based on the current fuel cycle parameters (e.g., steam pressure and flow, circulation ratio). Thus, the licensee has concluded that all tubes in the Yankee Rowe steam generators are acceptable for continued service and that no hardware modifications, preventive tube plugging, or other naasures are necessary to preclude fatigue crack initiation.
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3.0 CONCLUSION
The staff concludes that the actions taken by the licensee resolve the issues identified in Bulletin 88-02 and are, therefore, acceptable. Consistent with staff finding No. 11 in Reference 1, the above findings are subject to the development of administrative controls by the licensee to ensure that updated stress ratio and fatigue usage calculations are performed in the event of any significant changes to the steam generator operating parameters (e.g., steam pressure and flow. circulation ratio) relative to the reference parameters assumed in the Yankee Rowe analyses.
4.0 R_EFERENCE 1.
Safety Evaluation Report, " Evaluation of Westinghouse Methodology to Address Item C.2 of NRC Bulletin 88-02" which was transmitted to Westinghouse by letter dated October 2, 1989. NRC Accession No.
8910310013 2.
" Technical Evaluation Report of Tube Fatigue Properties Presented in WCAP-12079, ' Yankee Rowe Plant Evaluation for Tube Vibration Induced Fatigue,'" prepared for NRC staff by Viking Systems International, October 1989.