ML20196E477
| ML20196E477 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/08/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19295G792 | List: |
| References | |
| IEB-88-002, IEB-88-2, NUDOCS 8812120009 | |
| Download: ML20196E477 (3) | |
Text
fwo%'o l'ulTED STATES g
U NUCLEAR REGULATORY COMMISSION l
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E b.SHINGTON, D. C. 20655 SAFETY EVALUAi10N BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING BULLETIl4 88-02, "RAPIDLY PROPAGATING FATIGUE CRACKS IN STEAM GENERATOR TUBES' NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS NOS. 1 AND 2 DOCKET 5 NOS. 50-282 AND 50-33 INTRODUCTION By letters dated March 24 and April 12, 1988, Northern States Power Company (the licensee) sQmitted its response to NRC Bulletin 88-02, "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes".
Bulletin 88-02 requested that licensees for plants with Westiaghouse steam generators employing carbon steel support plates take certain actions (specified 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.
DISCUSSION The .censee reports that the Prairie Island Unit 1 and Unit 2 steam generators exhibit evidence of denting at the uppermost support clate.
Accordingly, items C.1 and C.2 of the bulletin are applicable to Prairie Island Units 1 and 2.
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 24, 1988, submittal.
This enhar,Jed 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 uf these actions.
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 its April 12, 1988, submittal which included Westinghouse reports WCAP-11787 (Proprietary Version) and WCAP-11788 (Non *roprietary Version),
"Prairie Island Units 1 and 2, Evaluation for Tube Vit: ration Induced Fatigue",
March, 1988.
These reports provide a detailed description of the analyses which were conducted to establish the susceptibility of the Prairie Island steam generator tubes to rapidly propagating fatigue cracks and to identify any needed corrective actions.
The NRC staff has previously reviewed the Westinghouse generic program.
The staff's Safety Evaluation of the Westinghouse generic program is incorporated as part of this bafety Evaluation (SE) as Attachment 1 (proprietary version) and (nun proprietary version).
As stated in the attachments, the staff has concluded that the Westinghouse program is an acceptable approach for resolving item C.2 of the bulletin.
The staff has further concluded that the WestinghoJse program, if properly implemented, will provide reasonable assurance against future failures of the kind which occurred at North Anna Unit 1.
The 8812120009 881203 PDR ADOCK 05000282
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C PDC
staff's review of WCAP-11787 indicates that the Westinghouse generic program has been fully implemented for Prairie Island Units 1 and 2.
Key elements of the Westinghouse program are sunnarized in Section 3.3.6 of Attachment 1 of this SE.
The analyses documented in WCAP-11787 show that all unsupported tubes in the Prairie Island steam generators satisfy the Westinghouse stress ratio criterion.
The fatigue usage factor for the most limiting tube is calculated to be 0.173 over the projected 40-year lifetime of Units 1 and 2.
Thus, the licensee has cc ~ ided that all tubes in the Prairie Isiand steam generators are acceptable to antinued service and that no hardware modifications, preventive tube piugging, or other measures are necessary to preclude fatigue crack initiation.
The Prairie Is'and analyseo conservatively assumed that all unsupported tubes were dented at the uppermost support plate.
In addition, the stress ratio and fabgue estimates were based on the assumption of a full mean stress effect (i.e.,
yield stress), consistent with staff finding No. 3 in Section 4 of Attachments 1 and 2.
The original antivibration bar (AVB) supports in the Prairie Island ster' generators were replaced with a modified design some time ago, although he staff does not know exactly when this was done.
Evaluation of the eddy current depth variations are within one tube pitch)y uniform insertion depths (i.e.,
data revealed the modified AVBs to have ver 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 Prairie Island tubes were determined to be small compared to peaking factors which exist for certain tubes at other plants (including the tube that failed at North Anna) which continue to employ the original AVB design. The Prairie Island Units 1 and 2 peaking factor assessment was based on air model tests which considered the modified AVB geometry and AVB deptn measurement uncertainties.
A The Prairie Island stability ratio analysis was perfomed with the FASTVIB computer code using thennal-hydraulic input from the 3-D ATH0S Code for a reference operating cycle (i.e., cycle 12 for Prairie Island Unit 1).
CONCLUS10],{
The licensee's analyses documented in WCAP-11787 fully resolve the issues identified in Bulletin 88-02 and are acceptable. These analyses indicate that there is reasonable assurance that rapidly propagating fatigue cracks of the type which occurred at NL~h Anna Unit I should not occur at Prairie Island Units 1 and 2.
These finuings aie subject to the development of administrative controls by the licensee to ensure that updated stress ;Atlo and fatigue usage calculations are performed in the event of any signi"unt hanges to the steam generator operating parameters (e.g., steam flow ant pe n re, circulation ratio) relative to the reference parameters assumed in the ~AP-11787 analyses.
Based on the satisfactory resolution of item C.2 of Bulletin 88-02, the licenseo may, at its option, terminate its commitment to the enhanced leak rate monitoring program described in the licensee's March 24. 1988 submittal.
Nevertheless, the
. o staff encourages the licensee to review its leak rate monitoring procedures to ensure the continued effectiveness of these procedures for the timely detection, monitoring, and trending of rapidly increasing leak rates.
Date:
Principal Contributor:
E. Hurphy Attachments:
1.
Proprietary SE 2.
Non-proprietary SE f
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