ML20213A021

Demonstration of Conformance of Prairie Island Units to App K & 10CFR50.46 for Large Break LOCAs (Fq Equals 2.36 & F- Delta-H Equals 1.63)

ML20213A021
Person / Time
Site:	Prairie Island
Issue date:	04/30/1987
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20212R565	List: ML20212R562 ML20212R601 ML20212R632 ML20212R689 ML20213A021 ML20213A047
References
TAC-65078, TAC-65079, NUDOCS 8704270345
Download: ML20213A021 (15)

Text

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Exhibit E Prairie Island Nuclear Generating Plant License Amendment Request Dated April 13, 1987 Demonstration of the Conformance of the Prairie Island Units to Appendix K and 10 CFR 50.46 for Large Break IDCAs (Fq - 2.36 and F-DELTA H - 1.63)

Prepared by: Nuclear Safety Department' Nuclear Technology Division Westinghouse Electric Corporation 45 87 427 DOCg05h202 ppg P

e DEMONSTRATION OF THE CONFORMANCE OF PRAIRIE ISLAND UNITS TO APPENDIX X AND 10CFR50.46 FOR LARGE BREAK IDCAs (Fg = 2.36 AND F-DELTA-H = 1.63) .

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Westinghouse Electric Corporation Nuclear Technology Division Nuclear Safety Department Safeguards Engineering and Development April 1987 i

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I. INTRODUCTION Thio document reports the results of an analysis which was performed to d:monstrate that Prairie Island Units 1 and 2 conform to the requirements of 10CFR50.46 (Reference 1) in accordance with Appendix K for.Large Break Lono-of-Coolant-Accidents (LOCA).

II. BACKGROUND Prior to the plant start-up of Cycle 11 of Prairie Island Unit 1, a large Brcak LOCA analysis was performed for Prairie Island Units 1 & 2 d0monstrating the conformance of these units to the Acceptance criteria of 10CFR50.46 through the and of cycle 11. The analysis was performed using the Westinghouse 1981 Evaluation Model (Reference 2) and resulted in a worst case (C D=0.4) peak clad temperature (PCT) of 2186 0F (including required penalties for Upper Planum Injection and transition coro hydraulic mismatch). The analysis accounts for filling of the guide thimbles during reflood and 5% Uniform Steam Generator Tube Plugging. In anticipation of the availability of a licensed Best Estimate Evaluation Mod 31 for UPI plants for cycle 12, the Cycle 11 81 EM analysis modeled ths Prairie Island core geometry to conservatively conform to the spscifics of the cycle 11 core. Specifically, the core was assumed to consist of one-third Westinghouse 14x14 Optimized (OFA) fuel and two-thirds ENC 14x14 "TOPROD" fuel. Inasmuch as the geometries of these fual types differ, the Cycle 12 core for Prairie Island, containing a higher proportion of OFA fuel, would not be adequately modeled by the Cycle 11 analysis. Therefore, an additional analysis ils required to i dsmonstrate the conformance of the Prairie Island units to 10CFR50.46 for Cycle 12.

l III. METHOD OF ANALYSIS Ths cycle 12 analysis assumed a core comprised of 84 assemblies of 14x14 OFA fuel and 37 assemblies of 14x14 "TOPROD" fuel. Since the fuel rod outer diameter for 14x14 OFA is smaller than that for 14x14 "TOPROD",

j cera flooding area is increased, decreasing core flooding rate and 4 providing a conservative modeling basis for cycle 12 l The analysis was performed using the Westinghouse 1981 Evaluation Model.

l Only the most limiting break case was re-analyzed since the cycle 11 analysis clearly demonstrated that the Cn=0.4 break is substantially

moro limiting than other discharge coefficients using the 81 EM for

, Prairie Island.

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i Tho Westinghouse 1981 ECCS Large Break Evaluation Model was developed to

datormine the RCS response to design basis large break LOCAs and consists

) of the SATAN-VI, HREFLOOD, COCO, and LOCTA IV computer codes (References i 3, 4, 5, and 6 respectively). The SATAN-VI code was used.to generate the i blowdown portion of the transient. The NREFLOOD code, which calculated the refill /reflood system hydraulics, is known to contain an dverly l connerygtive modeling of the Lower Plenum and Downconer metal heat

rolcase (Reference 7). The release of metal heat during the reflood transient is limited by conduction heat transfer, so that a solution of

{ tho conduction equation provides a realistic representation-of reflood l cotcl heat release. Such an approach has been found to be an acceptable i modoling of the release of metal heat during reflood (Reference 8).

NREFLOOD input values were adjusted to simulate the conduction limited 1 Ectal heat releases during reflood. The COCO code operates interactively with the NREFLOOD code to evaluate the containment pressure response.

, Cladding thermal analyses were performed with the LOCTA-IV code which l

unso the RCS pressure, fuel rod power history, steam flow past the

uncovered part of the core, and mixture height history from the SATAN-VI r

] cnd HREFLOOD codes as input. The hydraulic analyses and core thermal

! transient analyses assumed 102 percent of licensed NSSS core power. The

! analysis also assumed a hot channel enthalpy rise factor (F-DELTA-H) of 1.63 at a total peaking factor (F g) of 2.36.

! Tho Safety Evaluation Report on the Revised PAD Thermal Safety Model ,

l requires that an evaluation be performed to examine the effects of fuel l burnup on Peak Clad Temperature for steam cooling plants using the 1981 Evaluation Model with Revised PAD input. A study performed for Prairie l Island has demonstrated that Beginning-of-Life remains bounding in terms of Peak Clad Temperature throughout cycle 12.

i Table i shows the time sequence of events for the Large Break LOCA I transients. Table 2 provides a brief summary of the important results of

tho LOCA analyses for each case. Figures 1 and 2 show important core i l characteristics during the blowdown phase of the transient (Core Pressure i and Core Flow versus Time, respectively). Figures 3 and 4 indicate the

! flow of ECCS water into the RCS (Accumulator Flow and Pumped ECCS Flow l vorsus Time, respectively). The flooding rate during the reflood portion

, of the transient are given in Figure 5. Clad Average Temperatures as a function of time indicating peak clad temperatures are given in Figure

6. The Safety Injection (SI) system was assumed to be delivering to the RCS 22 seconds after the generation of a safety injection signal. The 22-second delay includes time required for diesel start-up and loading of tho safety injection pumps onto the emergency buses. Minimum safeguards Emorgency Core Cooling System capability and operability has, also, been cocumed in this analysis.

• The NRC has acknowledged this overconservatism by acceptance of WCAP-9561-P-A, Addendum 3 (Reference 7), finding it sufficiently compensatory to eliminate concerns regarding inadequate modeling of guide.

thimbles. Both the Cycle 11 and Cycle 12 analyses for Prairie Island have included specific and conservative modeling of guide thimbles.

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o III. RESULTS AND CONCLUSIONS Tha transient was considered to be terminated when the hot rod clad cvarage temperature " turned around" (i.e. - hot rod clad average temperature began to decline) indicating that the peak clad temperature htd been reached. The Cycle 12 analysis determined a peak clad tcmperature of 2086 0F. Current NRC restrictions require that a penalty b3 casessed and imposed for. insufficient modeling0 of upper plenum injcction. This penalty was assessed to be 054 F for the re-analyzed liniting case. In addition, a penalty of 10 F was imposed to account fcr hydraulic mismatch (crossflow) in the transition core. Imposition of thoce penalties results in a final peak clad temperature of 2150 F which is below the 22000 F Acceptance Criteria limit established by App ndix K of 10CFR50.46.

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i REFERENCES i

! 1. Bordelon, F. M., et al., LDCTA-IV Fresrent*Lons-of-Coolant l

Transient Analysis, WCAP 8301 (Proprietary Vergion), WCAP 8305 (Non-Proprietary Version), June 1974. >

2. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors: 10CFR 50.46 and Appendix K of 10CFR 50.46," Federal Register, Vol. 39,No. 3, i January 4, 1974. , ,

} 3. Bordelon, F. M., et al., BATAN-VI Proaram Cannrahansive j snace-Tina Danandant Analvais of Lons-of-coolant, WCAP 8302

{' (Proprietary Version) , WCAP 8306 (Non-Proprietary Version), Jun 1974.

4 I 4. Kelly, R. D., et al., Calculational Model for Core Refloodina after a Loss-of-coolant Accident (Wraflood Code), WCAP 8170 (Proprietary Version), WCAP 1871 (Non-Proprietary Version), Jun

! 1974.

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

t i 6. Eicheldinger, C., Westinahouse ECCS Evaluation Model, 1981 i j Version, WCAP 9220-P-A (Proprietary Version), WCAP 9221-A i (Non-Proprietay Version) , Rev. 1, 1981.

i I 7. Bordelon, F. M., H. W. Massie, and T. A. Zordan, Westinahouse ECCS Evaluation Model-Summarv, WCAP 8339, July 1974.

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l 8. Bordelon, F. M., et al., The Westinahouse ECCS Evaluation Model i Sunclementary Information, WCAP 8471 (Proprietary Version, WCAP i 8472 (Non-Proprietary Version), January 1975.

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l 9. Salvatori, R., Westinahouma ECCS - Plant Bensitivity Studies, j WCAP 8340 (Proprietary ~ Version) , WCAP 8356 (Non-Proprietary l

Version), July 1974.

10. Delsignore, T., et al., Westinahouma ECCS Two-Loon Sensitivity Studies (14 x 14), WCAP 8854 (Non-Proprietary Version),

September 1976.

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11. Westinahouse ECCS Evaluation Model Sensitivity Studies, WCAP 8341 (Proprietary Version) , WCAP 8342 (Non-Proprietary Version) , 1974. I

12. Kelly, R. D., C. M. Thompson, et al., Westinghouse Emergency Core '

Coolina System Evaluation Model for Analyzina Larne LOCAs Durina Operation With One Loon Out of Service for Plants Without Leon Isolation Valves, WCAP 9166, February 1978.

13. Eicheldinger, C., Westinahouse ECCS Evaluation Model. February 1978 Version, WCAP 9220 (Proprietary Version), WCAP 9221 (Non-Proprietary Version, February 1978.

14. Safety Evaluation Report on ECCS Evaluation Model for Westinahouse Two-Loon Plants, November 1977.

15. Letter from R. L. Kelly, Westinghouse Electric Corporation, to T.

R. Wilson, Wisconsin Electric Power Company, (WEP-78-2), dated February 24, 1978.

16. "NRC Questions Regarding the January 16, 1978 submittal by

. Westinghouse Designed Two-Loop Plant Operators," February 1, 1978.

17. Letter from T. M. Anderson, Westinghouse Electric Corporation, to  ;

J. Stolz, NRC, dated June 1978.

18. Letter from T. M. Anderson, Westinghouse Elestric Corporation, to J. Stolz, NRC, (NS-TMA-8130), dated June 20, 1978.

19. " Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants," March 1978.

20. Young, M. Y., " Addendum To: BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients (Special Report: Thimble Modeling in Westinghouse ECCS Evaluation Model)," WCAP-9561-P-A, Addendum 3 (Proprietary), 1986.

21. Young, M. Y., et al., "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9561-P-A (Proprietary), March 1984

.o TABLE 1 LARGE BREAK TIME SEQUENCE OF EVENTS DECIA (CD = 0.4)

(Sec)

Stort 0.0 R30ctor Trip 0.593 Signal S.I. Signal 0.64 Acc. Injection 9.29 End of Blowdown 22.693 Pump Injection 22.64 .

Ecttom of Core 35.567 R:covery Acc. Empty 44.33

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TABLE 2 LARGE BREAK ,.

RESULTS 3

DECLG (CD = 0.4)

Pe k Clad Temp., OF w/ penalties 2150.

Pe2k Clad Temp., OF 2086.

1 PeOk Clad Temp. 7.5 Location, ft Local Zr/ 6.371 H 2 O Rxn (max), 4 LocD1 Zr/ 7.25 H 2 O Location, ft i Total Zr/H 2 O Rxn, % <0.3 f Hot Rod Burst Time, 71.8 ccc H3t Rod Burst 7.25 Location, ft t

Calculation i

NSSS Power, MWt, 102% of 1650 Pack Linear Power, kw/ft, 102% of 14.62

PO2 king Factor (At Design Rating) 2.36 j Accumulator Water Volume (Cubic 4

Fcat per Tank-Nominal) 1270.

Accumulator Pressure, psia 700 i Nu;ber of Safety Injection Pumps Operating 2 l StOOm Generator Tubes Plugged 5%

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NOTE: Asterisks (*) do not represent a separate curve, but provide a tracer to

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CD=0.4 DECLG CLAD AVG. TEMP.HDT RCD BURST. 7.25 Fil ) PEAK, 7.50 FT(*)

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