ML20151Q220

Demonstration of Conformance of Prairie Island Units 1 & 2 to App K & 10CFR50.46 for Large Break Locas

ML20151Q220
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Site:	Prairie Island
Issue date:	08/31/1988
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20151Q212	List: ML20151Q208 ML20151Q220
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l DEMONSTRATION OF THE CONFORMANCE OF PRAIRIE IS1AND UNITS 1 AN', 9 TO APPENDIX X AND 10CFR% 44 FOR IARGE liREAX LOCAS Westinghouse Electric Corporation Nuclear Technology Systems Division Nuclear Safety Department Safeguards Engineering and Development August 1988 l

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I. Introduction This document reports the results.of an analysis that was performed to demonstrate that Prairie Island, Units 1 and 2, meet the requirements of Appendix K and 10CFR50.46 for Large Bre ak Loss-of-Coolant-Accidents (LOCA)

(Reference 1)

11. Method of Analysis The analysis was performed using the Westinghouse Large Break LOCA Best-Fstimate Methodology (References 2 and 3). The Westinghouse Best-Estimate Methodology was developed consistent with ,iidelines set forth in the SECY-83-472 document (Reference 4). These guidelines provide for the use nf realistic models and assumptions, with the exception of specific models and arsumptions required by Appendix K. The technical basis for the use of this model is discussed in detail in Reference 2 and 3.

The Best Estimate Methodology is comprised of the ECOBRA/ TRAC and COCO computer codes (References 3 and 5, respectively). The ECOBRA/ TRAC code was used to generate the complete transient (blowdown through reflood) system hydraulics as well as the cladding thermal analysis. The COCO code was used to generate the containment pressure response to the mass and energy release from the break. This containment pressure curve'was used )

as an input to the HCOBRA/ TRAC code.

The fuel parameters used as input for the LOCA analysis were generated using the Westinghouse fuel performance code (PAD 3.3, Reference 6). The fuel parameters input to the code were at beginni..g-of-life (maximum

, densification) values. In addition, a 10 F cladding tempers"ure penalty is applied to address mixed cores. This 10 F penalty is based on previous'EM model calculations.

In order to cocply with the Appendix K requirements, the code was prevented from returning to nucleate boiling after CHF during blowdown even if the flutd and rod surface conditions would permit this to occur.

Previous calculations indicated that the very end of the fuel rods, a low power and low temperature region, as well as most of the low power channel, would be calculated to return to nucleate boiling during blowdows. when the core flow would reverse at 10 seconds into the transient. The flow is sufficiently high, and of low void fraction, such that when these fluid conditions are combined with the lou surface temperatures, the code would calculate a nucleato boiling heat flux.

In order to comply with the Appendix K requirement of no return to nucleate boiling during blowdown, a "flag" (coding logic) was put into the code which limits the heat transfer coefficient to that of transition boiling when the code would have normally calculated a lower nuc1cate

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boilingheattransfarcoefficjent. The limited value of this heat transfer coefficient is 1800 Btu /hr-ft - F, which was taken from the blowdown heat transfer experiments conducted at Westinghouse, and is a value previously approved by the NRC Staff for the UHI model. The blowdown experiments and the basis for the value of the transition boiling heat transfer coefficient are given in Section 7 of Reference 3.

The analysis was performed using the four channel core model developed in Reference 3 for the 0.4 double-ended cold leg guillotine (DECLG) breaks.

A break size coefficient (CD) of 0.4 was found to be most limiting as documented in Reference 3. These transients were considered to be terminated if the hot rod cladding temperature began to decline and the

i. injected ECCS flows exceeded the break flow.

III. Results and Conclusions Table 1 shows the time sequence of events for the Large Break LOCA j transients. Table 2 provides a brief summary of the important results of l the LOCA analysis and shows compliance with the 10 CFR 50, Appendix K '

1 requirements. Figures 1 through 8 show important transient results for the limiting 0.4 DECLC break (four channel core model). Note on these figures that the break occurs at time 0.0. Figure 1 shows the core pressure during the transient. Figure 2 shows the vapor and liquid mass flowrate at the top of the hot assembly. Figures 3 and 4 show the collapsed liquid level in the downcomer and core hot assembly channel, respectively, indicating the refilling of the vessel. Figures 5 and 6 show the flow of ECCS water into the cold leg (accumulator and high head safety injection flow) with Figure 7 showing the flow of low head safety injection into the upper plenum (UPI flow) . Figure 8 shows the resulting peak cladding temperature for the 0.4 DECLC break as a function of time for each of the five fuel rods modeled. Rod 1 is the hot rod in the hot assembly channel, Rod 2 is the hot assembly average rod, Rods 3 and 4 represent average assemblies in the center of the core and Rod 5 represents the lower power assemblies at the edge of the core. The upper '

plenum injection system (RHR) was calculated to be delivering into the reactor at 19 seconds af ter the generation of a safety injection signal.

This delay includes the time required for developing full flow from the SI pumps. The analysis assumed reactor coolant pumps remain in operation in conjunction with no loss of offsite power. Sensitivity studies (Reference

3) show that continued operation of the reactor coolant pumps results in the worst peak cladding temperature. Minimum safeguards ECCS capability and operability has also been assumed.

No additional penalties were required for upper plenum injection since the Westinghouse Large Break LOCA Best- stimate Methodology models the RHR flow to be injected into the upper plenum. This analysis result is below the 2200 F Accepta~ a Criteria limit established by Appendix K of 10CFR50.46 (Reference 1).

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REFERENCES

1. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors: 10CFR50.46 and Appendix K of 10CFR50.46," Federal Register, Vol. 39, No. 3, January 4, 1974.

2. Hochreiter, L. E., Schwarz, W. R., Takeuchi, K., Tsai,'C. K.,.

and Young, M. Y., Westinchouse Larne-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, WCAP-10924-P, Vol. 1, (Proprietary Version), June, 1986.

3. Dederer, S. I., Hochreiter, L. E., Schwarz, W. R., Stucker, D.

L. , Tsai, C. K. , and Young, M. Y. , Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection, WCAP-10924-P, Vol. 2, Rev. 1, April, 1988.

4. NRC Staff Report, "Emergency Core Cooling System Analysis Methods," USNRC SECY-83-472, November, 1983.

5. Bordelon, F. M., and E. T. Murphy, Containment Pressure Analysis Code (COCO), WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), June, 1974.

6. Westinchouse Revised PAD Code Thermal ~ c,tv Model, WCAP-8720, Addendum 2 (Proprietary), and WCAP-8785 (Jon-Proprietary).

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TABLE 1 I

LARGE BREAK TIME SEQUENCE OF EVENTS Four Channel Core EVENT 0.4 DECLG (seconds) 1 Start 0.0 l Reactor Trip Signal 0.I Safety Injection (S.I.) 2.0 i Signal  !

High Head S.I. Begins 7.0 Blowdown PCT Occurs 7.5 Accumulator Injection 8.7 Low Head S.I. Begins 21.0 End of Bypass 25.1 Bottom of Core Recovery 35.0 Hot Rod Burst 30.3 Hot Assembly Average 38.4 Rod Burst Accumulator Water Empty 45.7 Accumulator Nitrogen 68.2 Injection Ends Reflood PCT Occurs 72.7

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TABLE 2 LARGE BREAK RESULTS Feur Channel Core EVENT 0.4 DECLG (seconds)

Peak Cladding Temp., OF 2031 + 10 = 2041 Peak Clad Temp. 7.0 Location, ft.

Local Zr/ Water Reaction 11.65 (max),%

Local Zr/ Water Reaction 4.66 Location ft.

Total Zr/ Water Reaction, % <0.3 Hot Rod Burst Time, sec. 30.28 Hot Rod Burst Location, ft. 4.66 Hot Assembly Burst 38.43 Time, sec.

Hot Assembly Eurst 5.25 Location, ft.

Hot Assembly % Blockage 30.53 j Calculation Input Values:

NSSS Power, Hwt, 102% of 1650.

Peak Linear Power, kW/ft, 102% of 15.789 Peaking Factor (At Design Rating) 2.50 Accumulator Water Volume 1270.

(Cubic ft. per tank, nominal)

Accumulator Pressure, psia 754.7 Number of Safety Injection Pumps 3 (0perating (1 RHR + 2 HHSI)

Steam Generator Tubes Plugged 10% j

PRESSURE (PSIA)

CHANNEL 10. NODE 7 TOP OF CORE

.25E+04-

.225E*04- A

- .2E 04-5 m .175E+04-S w .15E+04-c: >

a g .125E+04- I w

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1 750. 1 500. l 250.

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TIME (SECONDS 1 PLOT NO. 56 Figure 1. Core Pressure History

a LIQUID, VAPOR, AND ENTRAINED MASS FLOW TOP OF CORE - CHANNEL 12, NODE 7 (HOT ASSEMBLY) 1-LIQUID FLOW, 2-VAPOR FLOW, 3-ENTRAINED LIQUID FLOW 20.

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TIME (SECONDS) PLOT NO. 12 I

Figure 2. Liquid, Vapor, and Entrained Mass Flow Rate at Top of the Hot Assembly i

-2 -- - , - . . m. .e3 -. ..-... ..,m..

LIQUID LEVEL DO!JNCOt1ER LIQllID LEVEL

. CHANNELS 2,3,7,8,14,15,22,e3,29,30,37,38,43,4.4 50.

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TIME (SECONOS) PLOT NO. 69 Figure 3 Collapsed Liquid Level in the Downcomer

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LIQUID LEVEL CORE L10UID LEVEL CHANNEL 12. (HOT ASSEMBLY) 12.

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Figure 4 Collapsed Liquid Level in the Hot Assembly

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7 TOTAL FLOW ACCUMULATOR TO INTACT COLD LEG COMPONENT'10.~ CELL 3

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TIME (SECONDS) PLOT NO. 46 l

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Figure 5. Accumulator Flow to the Cold Leg

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TOTAL F L 0l4 HHSI TO INTACT COLD LEG COMPONENT 6 CELL 5 200.

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TIME (SECONDS) PLOT NO. 47 Figure 6. HHSI Flow to the Cold Leg

TOTAL FLOW RHR FLOW UPPER PLENUM COMPONENT 24, CELL 1 500.

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TIME (SECONDS) PLOT NO. 48 Figure 7. UPI Flow to the Upper Plenum _

CLADDING TEMPERATURE AT 7 FT ROD 1 -HOT ROD- CH 12. ROD 2 -HOT ASSEMBLY -CH 12 3-AVG ROD-CH 11. ROD 4-AVG ROD-CH 10. ROD 5-L.P. ROD -CH

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TIME ISECONDS) PLOT NO. I Figure 8. Cladding Temperature History at PCT 1.ocation

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