ML20147D963

Safety Evaluation Accepting Util 860404 Evaluation of Environ Qualification of Equipment Considering Superheat Effects of high-energy Line Breaks for Plants,Per IE Info Notice 84-90

ML20147D963
Person / Time
Site:	Wolf Creek, Callaway, 05000000
Issue date:	02/25/1988
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20147D942	List: ML20147D937 ML20147D963
References
IEIN-84-90, NUDOCS 8803040212
Download: ML20147D963 (3)

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SAFETY EVALUATION ,BY THE OFFICE OF NUCLEAR REACTOR REGULATION ENVIRONMENTAL EFFECTS OF MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT KANSAS GAS AND ELECTRIC COMPANY WOLF CREEK GENERATING STATION DOCKET NO. 50-482

1.0 INTRODUCTION

Py letters dated April 4, 1986, the SNUPPS utilities submitted a report entitled "Evaluation of Environmental Oualification of Equipment Considering Superheat. Effects of High Energy Line Breaks for Callaway Plant and Wolf Creek Generating Station." This report responded to the staff concerns described in IE Infomation Notice 84-90. For certain MSLB accidents, steam generator tube bundle uncovery occurs and may result in the release of superheated steam. This will raise the temperature in the break compart-ment and adjoining compartments to levels above those previously calculated.

The above report and a subsequent letter dated April 1,1987 from Wolf Creek Nuclear Operating Corporatien described an analysis of this condition which is applicable for both the Callaway Plant and Wolf Creek Generating Station.

2.0 EVALUATION In the analysis of rain steam line breaks (MSLB) with superheated steam, the mass and energy releases following a MSLB with superheated steam blowdown were obtained from Westinghouse report WCAP-10961 and were calculated using the Vestinghouse computer code LOFTPAN. The version of LOFTRAN used includes the modification to account for heat transfer to the steam during steam generator tube bundle uncovery. This mcdification is described in Westinghouse Topical Report WCAP-8860, Supplement 1, which the staff found acceptable in its safety evaluation transmitted to Westinghouse by letter dated May 27, 1986.

The SNgPPS util{ ties postuigted a spectrum of steam line break sizes (4.6 ft2 ,

1.0 ft , 0.7 ft and 0.5 ft ) at different power levels (102% and 70%) in thg main steam tunnel. The licensee fcund that a break size of less than 0.5 ft was less limiting than the larger breaks because it does not result in steam generator tube uncovery until after the operator response time for teminating auxiliary feedwater. The staff finds the postulated spectrum of breaks accept-able.

The SNUPPS utilities used the Bechtel computer code FLUD to calculate the com-partment environmental temperature profiles. Credit was taken for heat transfer to the concrete structures and structural steel. When the atmosphere was super-heated, a maximum of 8 percent of re-evapot-12ation ratio was assumed. The peak calculated pressure values in the break compartment were belov. the qualification 8803040212 080225 PDR ADOCK 05000482 G PDR

requirements used for safety-related equipment in the steam tunnel. However, the calculated peak environtrental temperature values exceeded the qualification limit of some affected safety-related components.

Sinc'e the component terrperatures generally lag the changes in the environ-mental temperature, SNUPPS perfomed a component themal lag analysis on certain components including main steam pressure transmitters, transmitter instrument cable, MSIV/MFIV solenoid valves, MSIV/MFIV wiring and lugs, MSIV/MFIV terminal blocks, MSIV/MFIV control cable, MSIV/MFIY limit switches, MSIV/MFIV conax connectors, MSIV/MFIV limit switch instrument cable, J-601A solenoid valve, and J-601A control cable. Four tires the maximum Uchida correlation and forced convection heat transfer coefficients (Hilpert correlation) were assumed in modeling the condensing mode and saturation mode respectively. The results of the utilities' calculation were presented in Table 3.4 (Rev. 1) of the April 1, 1987 letter.

The results showed that with a few exceptions, the equiprrent surface tempera-tures did not exceed the qualification temperature limits until the tirre when a Steam Line Isolation Signal (SLIS) or Feedwater Isolation Signal (FWIS) was initiated (from the begining of the accident to the time of the FWIS and/or SLIS is defined as the calculation tire). The exceptions are main steam pressure transmitter instrument cable, PSIV/MFIY limit switch instrument cable, and J-601A Control Cable which exceeded their correspcnding qualification temperature limits. ShUPPS provided failure modes and effects analysis (FMEA) to justify the above exceptions. The FPEA confimed that failure cf these components will not affect the ability to safely shutdown following the postulated steam line breaks as the equipment will either fail safe, or alternative capability is provided.

The staff has reviewed the SNUPPS' analysis and finds that, with one exception, the assumpticns and methodologies are consistent with the staff guidance in NUREG-0588, "Interim Staff Position or Environmental Qualification of Safety-Related Electrical Equipment." The exception is the method of calculating flow velocity for determining the forced convection heat transfer coefficient.

The applicant's velocity was obtained from dividing the volumetric break flow by the flow area, while in NUREG-0588 the velocity is a function of blowdown rate divided by the containment volume. The staff finds the SNUPPS' velocity calculation for determining the forced convection heat transfer coefficient to be apprcpriate, and therefore, acceptable for determining inter-compartmental flow.

In order to assess the SNUPPS' results, Battelle Pacific Northwest Laboratory (PNL) perforced an independent confirmatcry analysis for the staff using the modified COBRA-NC computer code. This code has previously been used by the staff for compartmental pressure /terrperature response analysis. The tredified cede incorporates a heat transfer model and re-evapcrization rrodel (8% re-evaporization rate) in accordance with Appendix B of NUREG-0588 and provides both the compartrent environmental tersperature and component themal lag analyses simultaneously. The resulting peak component temperatures from the confirmatory aralysis are consistently bounded by the SNUPPS' results.

In sumary, based on review of the assumptions and methodology of the SNUPPS' analyses for determining steam line break environments including superheated steam blowdown, and the results of the independent analyses, the staff concludes t'.at t the component ten:peratures calculated by SNUPPS are acceptable. In addition, by letter dated October 2,1987, Wolf Creek Nuclear Operating Corporation responded to a staff request for additional information. This response provided a discussien and explanation clarifying ccncerns involving the failure modes and effects analysis. As a result, based on the information provided in the submittals dated April a,1986, October 2,1987, and February IP.,1988 the staff finds the failure n. odes and effects analysis and calculation times acceptable.

3.0 CONr.LUSION Based on the staff review of the mass and energy release data, compartment temperature response analysis, component surface temperature analysis, failure modes and effects analysis, and the staff's independent analyses, the staff finds the licensee's evaluation of environmental qualification of safety-related equipment in the steam tunnel following a MSLB with superheated steam releases to be acceptable. Therefore, the staff concludes that proper environ-mer.tal qualification of safety-related equipment outside containment has been demonstrated.

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