ML20140A751
| ML20140A751 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/31/1985 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20140A716 | List: |
| References | |
| NUDOCS 8601230365 | |
| Download: ML20140A751 (18) | |
Text
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t EIHIBIT F PRAIRIE ISLAND NUCLEAR GENERATING PLANT License Amendment Request Dated January 13, 1986 2
DEMONSTRATION OF THE (DNFORMANG OF PRAIRIE ISLAND UNITS 10 APPENDIX K AND 10CFR50.46 FOR SMALL BREAK LOCAs W2stingnouse Electric Corporation Nuclear Technology Division Nuclear Safety Department Safeguards Engineering and Development October 1985 g601g{0 PDR PDR p.g P
i
I.
In*nc h tien This doceent reports the results of an analysis that was performed to demonstrate tnat Prairie Island, Units I and II, meet the requirements of Appendix K and 10CFR50.46 for small break LOCA. The analysis incorporates anticipated plant harmare modifications, i.e., the new upper reactor internals package and the tnimble plus removal, as well as increased levels of Fg and steam generator tube plugging.
II. Methed of Analysis The analysis was perfccmed using the ]d NOTRUMP and LOCTA computer codes for a spectra of small break sizes. These codes are incorprated in the approved Westingnouse ECCS Small Break Evaluation Model ceveloped to cetermine the RCS response to cesign basis small break LOCAs and to address the NRC concerns expressec in NUREG-0611, " Generic Evaluation of Feewater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants."
The NCIRUMP coc:puter code is a one-dimensional general network code consisting of a rA2::ber of SCVanced features, including the Calculation of thermal non-equilibrim in all fluid volmes, flow regime-depencent crif t flux calculations with counter-current flooding limitations, mixture lev'el tracking lo61c in multiple-stacked fluid noces, and regime-cepenoent heat transfer i
correlations. NOTRUMP includes the representation of the reactor core as heated control volmes with an associated bubble rise model to permit a transient mixture height calculation. The multinoce capability of the program enables an explicit and cetailed spatial representation of various system cocpanents. In particular, it enables a proper calculation of the benavior of tne loop seal curing a loss-of-coolant transient.
Claccing ther=al analyses are performed with tne LOCTA-IV (Reference 3) code whien uses the RC pressure, fuel rod power history, steam flow past the mcovered part of the core, and mixture height history from the N07 RUMP hycraulic calculations, as input.
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T Table 1 lists important input parameters and initial conditions med in the NOTRUMP analysis. The core power decay and axial power distribution are snown in Figures 1 and 2.
For tnese analyns, the SI delivery considers pmp injection flow which is depicted in Figure 3 as a ftnetion of RCS pressure.
This figure represents injection flow from the SI pmps based on performance curves degraced 5 percent from the design head. The Safety Injection (SI) system was assmed to be delivering to the RCS 25 seconds after the generation of a safety injection signal. The 25-second oelay includes time required for diesel startup and Icecing of the safety injection pumps onto the emergency buses. Minimm safeguarcs Emergency Core Cooling Systerr. capability and operability has, also, been assmed in this analysis.
The nycraulic ana.'.yses are performed with the NOTRUMP code using 102 percent of the licensec NSSS core pwer. The core thermal transient analyses are performed with the LOCTA-IV coce using 102 percent of licensed NSSS core pwer.
Three break size transients were evaluated, 3 inch, 4 inch, and 6 inch. These transients were considered to be terminated when the follming criteria were met:
- 1. RCS system pressure had decreased below the accumulator set pint and acetz:ulator flow hac been initiated.
- 2. The core had been recovered and the core / upper plentra mixture level was at or above the bottom of the vessel outlet nozzles.
- 3. The net flow to the RCS was psitive with accumulator and SI fim exceeding the break f1w.
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i III. Reste ts and cnne1 usi ons Of the tnree break sizes evaluated, core mcovery occurred only for the 4-inch break case, Figures 5, 7, and 10. Accumulator injection terminates the 3-inch break transient; producing a positive slope en the core mixture level curve. A similar result occurs for the 6-inch break. The 4-inch break shows two brief periods of ccre mcovery prior to accumulator 1rdection. During the second period of mcovery a PCT of 1000 F occurs, Figure 8.
This value is well below all Acceptance Criteria limits of 10CFR50.46 and is non-limiting in comparison to large break analysis results.
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REFERENCES i
1.
Lee, H., Rupprecht, S. D., Tauche, W. D., Schwar, W. R., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-100% P-1, August 1985.
l 2.
Meyer, P. E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10074-P-1, August 1985.
3 Bordelon, F. M.,8101 (Proprietary), and WCAP-8306 (Non-Proprietary), June et al., "LOCIA-IV Program: Loss-of-Coolant Transient Aralysis," WCAF-1974.
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TABLE 1 INPJT PAR AMETrR9 flSFB IN THE SMAf f BRCAV ANif YRFR Parameter 4m11 Brente Peak Linear Power (kw/ft) 15.03 (incluces 1025 factor)
Total Peaking Factor, F 2.50 g
Power Shape See Figure 2 Fuel Assembly Array 14x14 0FA Nominal Cold Les Accynulator 1266 Water Volme (ft"/accmulator)
Nominal Colc Leg Aeg/ accumulator) mulator 2000 Tank Volune (ft Minimun Colc Leg Accumulator 715 Gas Pressure (psia)
Pucped Safety ledection Flow See Figure 3 Steam Generator Initial Pressure (psia) 733.0 Steam Generator Tuoe Plugging Level %
10 Fuel Assemcly Thictle Flugs Removed Reactor Upper Internals Package New Design I
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TABLE 2 SM1? f RRFAY LOCA TIME MFOfffMCF OF EVFRE i
3 in. Break 4 in. Break 6 in. Break (sec)
(sec)
(sec)
Start 0.0 0.0 0.0 Reactor Trip 4.5 2.8 17 Top of Core Uncovered WA 183 7 WA Cold Leg Accumulator Injection 703.0 371.5 162.6 Peax Cad Temperature Occurs WA 297 8 WA Tcp of Core Covered WA 325.8 WA i
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TABLE 3 SMALL BREAK LOCA RESULE FUEL C_ADDIE DATA 3 in. Break 4 in. Break 6 in. Break 4
Results Peak Cad Temperature ( F)
WA 1000 W/.
Peak Cad Location (ft)
WA 12.00 WA t
Local Zr/P 0 Resction (max), (5)
WA 0.066 WA 2
Local Zr/P 0 Reaction Location (ft)
WA 12.00 WA 2
Total Zr/H O Reaction (%)
<.3
<.3
<.3 2
Hot Roc Burst Time (sec)
WA WA WA Hot Rod Burst Location (ft)
WA WA WA 4
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IC' TIME AFTER SHUTCOwN ISECONCS) i l
9 Figure 1. Core Power After Reactor Trip F-9
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Figure 2. Small Break Power Distribution Assumed for Loca Analysis J
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Figure 3. Safety injection Flowrate Versus Pressure F-11
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1 Figure 4. 3 Inch Cold Leg Break RCS Pressure Versus Time f
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Figure 5. 3-inch Cold Leg Break Core Mixture Level Versus Time F-13
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Figure 6. 4 inch Cold Leg Break RCS Pressure Versus Time F-14
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Figure 7. 4 Inch Cold Leg Break Core Mixture Level Versus Time
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Figure 9. 6 Inch Cold Leg Broek RCS Pressure Versus Time F-17
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