ML20236U346

Submits Response to NRC 980326 RAI Re Computer Analysis Performed for Catawba Unit 1 Replacement Recirculating SG W/Respect to Low Temp Overpressure Protection

ML20236U346
Person / Time
Site:	Catawba
Issue date:	07/22/1998
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M99629, TAC-M99630, NUDOCS 9807300066
Download: ML20236U346 (8)

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• Duks Power Catawba Nuclear Station

, 4800 Cono rd Road York SC 29745 Cary R. Peterson I##3) #3#'DI "

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July 22, 1998 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20535

Subject:

Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Reoly to NRC Request for Additional Information on Amendment Request Regarding Pressure / Temperature Limits TAC Numbers M99629 and M99630

References:

1. Letter from Gary R. Peterson, Duke, to NRC, same subject, dated April 27, 1998

2. Letter from Peter S. Tam, NRC, to Gary R.

Peterson, Duke, same subject, dated March 1 26, 1998 l In Reference 1, Duke Energy Corporation responded to Questions 3 and 4 of the request for additional information contained in Reference 2. Pursuant to 10 CFR 50.4 and 10 CFR 50.90, Duke Energy Corporation is hereby providing additional information in support of the Reference 1 response.

This additional information concerns computer analysis performed for the Catawba Unit 1 replacement Recirculating g'/

Steam Generators (RSGs) with respect to the low temperature //

overpressure protection (LTOP) heat input transient. The additional information is contained in Attachment 1 to this #

' letter.

If you have any questions concerning this information, TWl l please call L.J. Ruay at (803) 831-3084.

9807300066 990722 7 PDR ADOCK 05000413 P PDR :-

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U.S. Nuclear Regulatory Commission July 22,fl998

.Page 2 Very truly yours, j/hk L Gary R. Peterson LJR/s i

Attachment J

xc (with attachment): l L.A. Reyes U.S. Nuclear Regulatory Commission Regional Administrator, Region II f Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 D.J. Roberts Senior Resident Inspector (CNS)

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U.S. Nuclear Regulatory Commission I Catawba Nuclear Station P.S. Tam NRC Senior Project Mannger (CNS)

U.S. Nuclear Regularcry Commission Mail Stop O-14H25  ;

Washington, D.C. 20555-0001 I i

M. Batavia, Chief I

-Bureau of Radiological Health I 2G00 Bull St. l Columbia, SC 29207 I l

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. 1 ATTACIDENT 1 Note: All references listed in this attachment are located on pages 4 and 5 of this attachment.

This additional information is being provided to the NRC concerning computer analysis recently conducted by Duke Energy Corporation regarding the heat input tr=nsient for the Catawba Unit 1 replacement steam geLerators (RSGs). Specifically, the steam generators in Catawba Unit 1 were recently replaced, increasing the heat transfer ar(a available for reverse heat transfer. The increased heat transfer area could increase the pressure response for the LTOP heat input transient originally documented in Reference 1. The information that follows summarizes the recently performed Duke Power RSG-specific analysis inputs and results.

The most limiting LTOP heat input transient involves the startup of a sin w le reactor coolant pump from a low temperature, water solid condition. The steam generator secondary is assumed to be 50 F warmer than the bulk temperature of the RCS. The increase in RCS flow results in a heatup and pressurization of the RCS.

The RETRAN-02 MODS .1 code (Reference 2) is utilized for the analysis. The SER limitations associated with the use of the RETRAN code are documented in References 3, 4 and 5, and have been reviewed for applicability to the simulation of the LTOP heat input transient. Two of the SER restrictions can be considered to apply to the LTOP heat input analysis. Reference 3, Finding 3 states the following in relation to the pressurizer:

"The nonequilibrium pressurizer model has not been qualified for those situations where the pressurizer becomes completely empty or full (goes from two phase to single phase).

Appropriate qualification should be submitted with any analysis that places the pressurizer in either of these modes."

The LTOP scenario of interest models the pressurizer as initially water solid, and remaining water solid for the duration of the transient. Under this situation, a

, single phase is maintained in the pressurizer, and RETRAN essentially treats the pressurizer as a normal l HEM volume. The above restriction is therefore not applicable.

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Another potential restriction concerns Reference 3,

. Finding 5:

"The use of the transport delay option should be limited to those cases showing a dominant flow direction."

Although the LTOP scenario of interest involves reverse flow in the idle loops, the flow direction does not change once established. Therefore, dominant flow directions are maintained,. and this restriction is therefore also not applicable.

The RETRAN SER restrictions have therefore been reviewed for applicability to the LTOP analysis. The restrictions listed in References 3, 4 and 5 do not restrict the application of the RETRAN code to perform the LTOP heat input transient. I The NRC-approved RETRAN RCS model for Catawba is described in Reference 6. A two-loop RETRAN model with cne single loop and one triple loop is selected in order to capture the flow asymmetry. The standard model is used with one significant exception. Since the transient initiates at zero power, a single volume SG secondary model is utilized. The single volume uses the bubble rise option with the local conditions heat transfer model applied to the steam generator tube

, conductors. The SG secondary side tube heat conductor heat transfer coefficients are determined using the McAdams correlation (free convection), which is consistent with the Reference 1 analysis. Zero tube plugging is modeled to maximize the reverse heat transfer. With this modeling approach, the low temperature, zero power conditions can be obtained, and the secondary-to-primary heat transfer can be simulated. I The transient begins with the rapid startup of a reactor coolant pump from a water solid, stagnant flow condition. The RCP is assumed to reach full speed in 9 seconds. Initial RCS temperatures of 250, 180 and 100 0F are analyzed. The corresponding SG temperatures are 300, 230 and 150 oF, respectively. The initial indicated RCS pressure is assumed to be 300 psig, as measured by the ', tide range RCS pressure taps on the hot legs. The actual RCS pressure is 355 psig, which accounts for WR pressure indication uncertainties. An initial SG mass of 215,000 lbm is assumed, which is a conservatively high initial mass.

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1 A single pressurizer PORV is credited to provide

. pressure relief. The PORV is actuated by the indicated WR ' RCS pressure signal. The LTOP PORV setpoint is assumed to be 400 psig, with a conservatively slow total opening delay of 3 seconds. The PORV reseat setpoint is assumed to be 370 psig, with a conservatively fast total closing delay of 1 second. The PORV critical flow contraction coefficient is adjusted to give a relief flow of 90% of the best-estimate relief capacity based on valve test data.

The results of the system pressurization are shown in Figure 1 (reactor vessel downcomer pressure). The downcomer volume pressure is measured 19.66 ft above the lowest point in the bottom of the reactor vessel. The pressure difference between this location and the bottom of the reactor vessel is less than 10 psi. For the 250 and 180 F cases, the peak RCS pressure occurs on the second cycling of the PORV. This is a result of using a faster total PORV closing delay.

The peak pressure results are also listed in Table 1 where they are compared to the proposed pressure-temperature limits taken from Reference 7. The peak pressure for each case is well below the pressure limit allowed for the initial operating temperature. It can therefore be concluded that the LTOP mitigation system successfully limits the magnitude of RCS pressure for the limiting heat input transient with replacement steam ,

generators. These results also confirm that the i limiting LTOP events for Catawba Unit 1 remain those that involve the mass input transients.

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Table 1 I

,- ,CNS Unit 1RSG LTOP Heat Input Transient Results Initial Temperatures Peak RV Beltline P-T Limit RCS / SG Pressure (psig)

( F) (psia / psig) 250 / 300 575.6 / 560.9 > 1000 180 / 230 587.3 / 572.6 > 1000 100 / 150 '526.3 / 511.6 621 References

1. " Pressure Mitigating Systems Transient Analysis Results," Westinghouse Electric Corporation, Utility Group on Reactor Coolant System Overpressurization, July 1977.

2. "RETRAN-02: A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," EPRI NP-1850-CCM, Revision 4, EPRI, November, 1988.

3. Letter frcm Cecil O. Thomas (USNRC) to Dr. Thomas W.

Schnatz (UGRA), " Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, 'RETRAN - A Program for One Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,' and EPRI NP-1850-CCM,

'RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,'" September 2, 1984.

4. Letter from Ashok Thadani (USNRC) to R. Furia (GPU),

" Acceptance for Referencing Topical Report EPRI-NP-1850 CCM-A, Revisions 2 and 3 Regarding RETRAN02/ MOD 003 and MOD 004," October 19, 1988.

5. Letter from Ashok Thadani (USNRC) to W. James Boatwright (TUEC), " Acceptance for Use of RETRAN02 MODS.0,"

November 1, 1991.

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6. DPC-NE-3000-PA, " Thermal-Hydraulic Transient Analysis Methodology," Duke Power Company, Revision 1, December

• 1997.

7. WCAP-13720, " Analysis of Capsule Y from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program," June 1993.

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