ML20246H012

Safety Evaluation Accepting Util 880325 & 1117 Responses to NRC Bulletin 88-002, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes

ML20246H012
Person / Time
Site:	Point Beach
Issue date:	07/07/1989
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20246G999	List: ML20246G995 ML20246H012
References
IEB-88-002, IEB-88-2, NUDOCS 8907140299
Download: ML20246H012 (3)

Text

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FELAlED TO CLOSEOUT OF BULIEilN 88-02 FACILITY OPERATING LICE!Gt NO. DPR-27 WISCONSIN ELECTRIC POWER COMPANY POIhl BEACH NUCLEAR PLANT, UhlT NO. 2 DOCKET NO. 50-301 INTRODUCTION By letters dated March 25 and November 17, 1988, Wisconsin Electric Power Company (the licensee) submitted its response to NRC Bulletin 88-02, " Rapidly Propagating Fatigue Cracks in Steam Generator Tubes." Bulletin 88-02 requested that licensees.for plants with Westinghouse steam generators employing carbon steel support plates take certain actions (specified in the bulletin) to ninimize 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 Point Beach Unit I steam generators were replaced in 1983 with a design that included stainless steel tube support plates. As a result, Bulletin 88-02 does not apply to Point Beach Unit 1. The licensee reports that the Point Beach Unit 2 steam generators exhibit evidence of denting at the uppermost support plates. Accordingly, items C.1 and C.2 of the bulletin are applicable to Point Beach Unit 2.

In accordance with item C.1 of the bulletin, the licensee has implemented an enhanced primary-to-secondary leak rate monitoring program. This enhanced leak rate monitoring program is an interim compensatory measure pending comple-tion of the actions requested in iten C.2 of the bulletin and NRC staff review a d approval of these actions. This prograre is intended to ensure the leakage from a rapidly propagating fatigue crack will be detected and monitored in a timely manner to ensure that the plant is shut down before rupture ocurs.

The licensee has implemented the generic program developed by Westinghouse to resolve item C.2 of the bulletin. The licensee's implementation of this i program was initially described in Westinghouse reports WCAP-11666 (proprietary version) and WCAP-11667 (non-proprietary version), " Point Beach Unit 2, Evalua-tion for Tube Vibration Induced fatigue." These reports provide a detailed description of the analyses which were conducted to estrblish the susceptibility of the Point Beach Unit 2 steam generator tubes to rapid 4 propagating fatigue cracks and to identify any needed corrective actions.

The Point Beach Unit 2 analyses conservatively assumed that all unsupported tubes were dented at the uppermost support plate. In addition, the stress ratio and fatigue estimates were based on the assumption of the full mean

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELAlED TO CLOSE0VT OF BULLETIN 88-02 FACILITY OPERATING LICENSE NO. DPR-27 WISCONSIN ELECTRIC POWER COMPANY POINI BEACH NUCLEAR PLANT, UHll NO. 2 DoctET NO. 50-301 INTRODUCTION j By letters dated March 25 and November 17, 1988, Wisconsin Electric Power Company (the licensee) submitted its response to NRC Bulletin 88-02, " Rapidly Propagating Fatigue Cracks in Steam Generator Tubes." Sulletin 88-02 requested that licensees for plants with Westinghouse steam generators employing carbon steel support plates take certain actions (specified in the bulletin) to ninimize 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 1 July 15,1987.

DISCUSSICN Point Beach Unit I steam generators were replaced in 1983 with a design that included stainless steel tube support plates. As a result, Bulletin 88-02 does not apply to Point Beath Unit 1. The licensee reports that the Point Beach Unit 2. steam generators exhibit evidence of denting at the uppermost support plates. Accordingly, items C.1 and C.2 of the bulletin are applicable to Point Beach Unit 2.

In accordance with item C.1 of the bulletin, the licensee has implemented an enhanced primary-to-secondary leak rate n.onitoring program. This enhanced leak rate monitoring program is an interim compensatory measure pending comple-tion of the actions requested in iten C.2 of the bt11etin and NRC staff review and approval of'these actions. This program is intended to ensure the leakage from a rapidly propagating fatigue crack will be detected and monitored in a timely manner to ensure that the plant is shut down before rupture occurs.

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 was initially described in Westinghouse reports WCAP-11666 (proprietary version) and WCAP-11667 (non-proprietary version), " Point Beach Unit 2, Evalua-tion for Tube Vibration Induced fatigue." These reports provide a detailed description of the analyses which were conducted to establish the susceptibility of the Point Beach Unit 2 steam generator tubes to rapidly propagating fatigue cracks and to identify any needed corrective actions.

The Point Beach Unit 2 analyses conservatively assumed that all unsupported tubes were dented at the uppermost support plate. In addition, the stress ratio and fatigue estimates were based on the assumption of the full mean l 8907140299 890707 .

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stress effect (i.e., yield stress). A fluid-elastic stability ratio analysis was performed with the FASTVIB computer code using thermal-hydraulic input from the 3-D ATH0S code for a reference operating cycle (i.e., cycle 13). The stability ratio estimates were adjusted to reflect flow peaking effects associated with actual antivibration bar (AVB) insertion patterns in the Point Beach Unit 2 steam generators. The flow peaking factors were estimated on the basis of preliminary data from Westinghouse air model tests.

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The Westinghouse generic program has undergone a number of refinements since issuance of WCAP-11666 and WCAP-11667 for Point Beach Unit 2. These refinements have included more accurate methods for determining AVB insertion depths, con-sideration of revised and additional data from air model tests concerning ,

flow peaking effects associated with specific AVB insertion depth configurations,  !

and consideration of uncertainties associated with AVB insertion depth estimates  ;

and air model tests data (

Reference:

Westinghouse report WCAP-11799 (proprietary l version) and WCAP-11800 (non-proprietary version), " Beaver Valley Units 1 and 2 -

Evaluation for Tube Vibration Included Fatigue," April 1988, NRC Accession No. 8805160073). Generic aspects of the refined Westinghouse program in WCAP-11799 were reviewed by the staff and found to be an acceptable approach for resolving item C.2 of the bulletin (

Reference:

Memorandum dated August 19, 1988; C. Y. Cheng to D. Neighbors, enclosing staff's safety evaluation of the Westinghouse generic program).

During a phone conversation on October 25, 1988, the staff requested that the licensee critically review the Point Beach analysis in WCAP-11666 in light of the new and revised information and data contained in more recent Westinghouse reports (e.g., WCAP-11799). As part of this request, the licensee was requested to respond to several questions concerning AVB measurement uncertainties and flow peaking factors. Formal documentation of this request wcs provided to the licensee by letter dated November 14, 1988.

The licensee provided a detailed response to the staff's request by letter dated November 17, 1988, transmitting WCAP-12043 (non-proprietary version)g ,

WCAP-12042

" Response (proprietary to NRC Questions." The response version) and included a refined assessment of AVB insertion depths including consideration of" measurement errors. Flow peaking factor assumptions were updated to reflect the revised AVB assessment and the. latest Westinghouse flow peaking data.

Stability ratio and stress ratio estimates were updated accordingly and compared to the westinghouse stability ratio and stress ratic acceptance criteria which have been previously reviewed and approved by the NRC staff (see above mentioned NRC staff memorandum dated August 19,1988).

The updated analyses demonstrate that all unsupported tubes in the Point Beach Unit 2 steam generators are well within the stability ratio and stress ratio  !

acceptance criteria, and thus will not be subject to fatigue crack initiation over the projected lifetime of the plant. Thus, the licensee has concluded that all tubes in the Point Beach Unit 2 steam generators are acceptable for continued service and that no hardware modifications, preventive tube plugging, or other measures are necessary to preclude fatigue crack initiation.

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J CONCLUSION The licensee's analyses fully resolve the issue 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 North Anna Unit I should not occur at Point Beach Unit 2.

These findings are subject to the development of administrative controls by the licensee to ensure that updated stress ratio and fatigue usage calculations are perforned in the event of any significant changes to the steam generator operating parameters (e.g., steam flow and pressure, circulation ratio) that are nonconservative relative to the reference parameters assumed in the licensee's analyses.

Based on the satisfactory resolution of item C.2 of Bulletin 88-02, the licensee may, at its option, terminate its commitment to the enhanced leak rate monitoring program described in its November 11 and November 20, 1987 and March 25, 1988 submittals. Nevertheless, the 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.

Principal Contributor: E. Murphy Dated: July 7, 1989

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