ML20213A047
| ML20213A047 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/30/1987 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20212R565 | List: |
| References | |
| TAC-65078, TAC-65079, NUDOCS 8704270360 | |
| Download: ML20213A047 (15) | |
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Exhibit F 1
Prairie Island Nuclear Generating Plant License Amendment Request Dated April 13, 1987 I
t Demonstrstion of the Conformance of the Prairie Island Units
- to Appendix K and 10 CFR 50.46 for Large Break thCAs j (Fq - 2.32 and F-DELTA-H - 1.66). ;
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j r y Prepared by: Nuclear Safety Department l Nuclear Technology Division :
) Westinghouse Electric Corporation '
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- 0704270360 870413 PDR ADOCK 05000282 l
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DEMONSTRATION OF THE CONFORMANCE OF PRAIRIE ISLAND UNITS 1
! TO
' APPENDIX K AND 10CFR50.46 1
FOR LARGE BREAK LOCAs j
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Westinghouse Electric Corporation i Nuclear Technology Division I
Nuclear Safety Department Safeguards Engineering and Development !
l April 1987 l
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l I. INTRODUCTION i This document reports the results of an analysis which was. performed to l
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 l Loco-of-Coolant-Accidents (LOCA).
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II. BACKGROUND 1
I Prior to the plant start-up of Cycle 11 of Prairie Island Unit 1, a large Brask LOCA analysis was performed for Prairie Island Units 1 & 2 i dsmonstrating the conformance of these unitsThe to the Acceptance analysis was performed criteria.
of 10CFR50.46 through the and of Cycle 11.
using the Westinghouse 1981 Evaluation Model (Reference 2) and 0 resulted in a worst case (C D=0.4) peak clad temperature (PCT) of 2186 F (including required penalties for Upper Planum Injection and transition core 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 Model for UPI plants for cycle 12, the Cycle 11 81'EM analysis modeled tho Prairie Island core geometry to conservatively conform to the cpacifics of the cycle 11 core. Specifically, the core was fuel and assumed to l
consist of one-third Westinghouse 14x14 Optimized Inasmuch as(OFA) the geometries of these two-thirds ENC 14x14 "TOPROD" fuel.
i-funl 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 is required to I
- dsmonstrate the conformance of the Prairie Island units to 10CFR50.46 for Cycle 12.
1 III. METHOD OF ANALYSIS l
The Cycle 12 analysis assumed a core comprised of 84Since assemblies the fuel of 14x14 rod OFA fuel and 37 assemblies of 14x14 "TOPROD" fuel. l cuter diameter for 14x14 OFA is smaller than that for 14x14 "TOPROD",
core flooding area is increased, decreasing core flooding rate and providing a conservative modeling basis for Cycle 12 1 1
The analysis was performed using the Westinghouse 1981 Evaluation Model.
Only the most limiting break case was re-analyzed since the Cycle 11 analysis clearly demonstrated that the Cn=0.4 break is substantially more limiting than other discharge coefficients using the 81 EM for Prairie Island.
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! The Westinghouse 1981 ECCS Large Break Evaluation Model was developed to datermine the RCS response to design basis large break LOCAs and consists i of the SATAN-VI, HREFLOOD, COCO, and LOCTA IV computer codes (References :
l The SATAN-VI code was used to generate the 3, 4, 5, and 6 respectively).
l blowdown portion of the transient. The HREFLOOD code, which calculated the refill /reflood system hydraulics, is known to contain an overly t
conserygtive modeling of the Lower Plenum and Downcomer metal heatThe rele release (Referenco 7).
transient is limited by conduction heat transfer, so that a solution of tha conduction equation provides a realistic representation of reflood matal heat release. Such an approach has been found to be an acceptable modeling of the release of metal heat during reflood (Reference 8).
EREFLOOD input values were adjusted to simulate the conduction limitedThe CO ;
antal heat releases during reflood.
with the EREFLOOD code to evaluate the containment pressure response.
Cladding thermal analyses aere performed with the LOCTA-IV code which unas the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the SATAN-VI I and NREFLOOD codes as input. The hydraulic analyses and core thermal The transient analyses assumed 102 percent of licensed NSSS core power. of analysis also assumed a hot channel enthalpy rise factor (F-DELTA-H) 1.66 at a total peaking factor (Fg ) of 2.32.
Tha Safety Evaluation Report on the Revised PAD Thermal Safety Model
- requires that an evaluation be performed to examine the effects of fuel burnup on Peak Clad Temperature for steamAcooling plants usingforthe study performed 1981 Prairie Evaluation Model with Revised PAD input.
Island has demonstrated that Beginning-Of-Life remains bounding in terms of Peak Clad Temperature throughout Cycle 12.
e Table 1 shows the time sequence of events for the Large Break LOCA transients. Table 2 provides a brief summary of the important results of J the LOCA analyses for each case. Figures 1 and 2 show important core characteristics during the blowdown phase of the transient (Core Pressure and Core Flow versus Time, respectively). Figures 3 and 4 indicate the l
(Accumulator Flow and Pumped ECCS Flow flow of ECCS water into the RCS The flooding rate during the reflood portion I
versus Time, respectively).
of the transient are given in Figure 5. Clad Average Temperatures as a l function of time indicating peak clad temperatures are given in Figure The Safety Injection (SI) system was assumed to be delivering to the 6.
RCS 22 seconds after the generation of a safety injection signal. The 22-second delay includes time required for diesel start-up andsafeguards Minimum loading of the safety injection pumps onto the emergency buses.
' Emergency Core Cooling System capability and operability has, also, been assumed 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 l
i have included specific and conservative modeling of guide thimbles.
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III. RESULTS AND CONCLUSIONS Tha transient was considered to be terminated (i.e. - hot when rod cladtheaverage hot rod clad svarage temperature temperature began to" turned around" indicating that the peak clad temperature decline) hr.d been reached. 0 The cycle 12 analysis determined a peak clad tcmperature of 2061 F. Current NRC restrictions require that a penalty b2 assessed and imposed for insufficient modeling 0 of upper plenum injcction. This penalty was assessed to be 054 F for the re-analyzed limiting case. In addition, a penalty of 10 F was imposed toImposition account of for hydraulic mismatch (crossflow) in the transition core. 0
- thsce penalties results 0in a final peak clad temperature of 2125 F which is below the 2200 F Acceptance Criteria limit established by App 2ndix K of 10CFR50.46.
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l REFERENCES )
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- 1. Bordelon, F. M., et al., LOCTA-IV Proaram: Loss-of-Coolant Transient Analvsis, WCAP 8301 (Proprietary Version), WCAP i 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 Resister, Vol.
39,No. 3, January 4, 1974.
- 3. Bordelon, F. M., et al., SATAN-VI Proaram: Comorehensive Space-Time Denendent Analysis of Loss-of-Coolant, WCAP 8302 (Proprietary Version), WCAP 8306 (Non-Proprietary Version), June 1974.
- 4. Kelly, R. D., et al., Calculational Model for Core Refloodina after a Loss-of-Coolant Accident (Wreflood code), WCAP 8170 (Proprietary Version) , 'n' CAP 1871 (Non-Proprietary Version), June 1974.
- 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. Eicheldinger, C., Westinahouse ECCS Evaluation Model. 1981 Version, WCAP 9220-P-A (Proprietary Version), WCAP 9221-A (Non-Proprietay Version) , Rev.1, 1981. ,
- 7. Bordelon, F. M., H. W. Massie, and T. A. Zordan, Westinahouse ECCS Evaluation Model-Summarv, WCAP 8339, July 1974.
- 8. Bordelon, F. M., et al., The Westinchouse ECCS Evaluation Model: Sucolementary Information, WCAP 8471 (Proprietary Version, WCAP 8472 (Non-Proprietary Version), January 1975.
- 9. Salvatori, R., Westinahouse ECCS - Plant Sensitivity Studies, WCAP 8340 (Proprietary Version) , WCAP 8356 (Non-Proprietary Version) , July 1974.
- 10. Delsignore, T., et al., Westinchouse ECCS Two-Loon Sensitivity Studies (14 x 14), WCAP 8854 (Non-Proprietary Version), September 1976. ,
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(Proprietary Version) , WCAP 8342 (Non-Proprietary Version), 1974.
- 12. Kelly, R. D., C. M. Thompson, et al., Westinahouse' Emeraency Core Coolina System Evaluation Model for Analyzina Larae LOCAs Durina Operation With One Loon Out of Servica for Plants Without Loon 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 Reoort on ECCS Evaluation Model for Westinchouse Two-Loon Plants, November 1977.
- 15. Letter from R. L. Kelly, Westinghouse Electric Corporation, to T.
R. Wilson, Wisconsin Electric Power Company, (NEP-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. Stciz, 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 l
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TABLE 1 I LARGE BREAK .
TIME SEQUENCE OF EVENTS DECIG (CD = 0.4)
(Sec) 0.0 Start 0.592 Rasctor Trip Signal O.63 S.I. Signal 9.30 Acc. Injection 22.696 End of Blowdown 22.63 Pump Injection 35.56 Bottom of Core Recovery 44.326 Acc. Empty
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TABLE 2 LARGE BREAK RESULTS .
DECIA (CD = 0.4)
Psak Clad Temp., OF 2125.
w/ penalties 2061.
Pack Clad Temp., OF 7.5 P32k Clad Temp.
Location, ft 5.623 Local Zr/
H 2 O Rxn (max), %
7.25 Local Zr/
H 2 O Location, ft
<0.3 Total Zr/H 2 O Rxn, %
Hot Rod Burst Time, 73.2 sec ,
7.25 Hot Rod Burst Location, ft Calculation:
NSSS Power, MWt, 102% of 1650 Paak Linear Power, kw/ft, 102% of 14.37 Paaking Factor (At Design Rating) 2.32 Accumulator Water Volume (Cubic 1270.
Feet per Tank-Nominal) 700 Accumulator Pressure, psia l Number of Safety Injection Pumps i 2
Operating 5%
Steam Generator Tubes Plugged l 1
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CD=0.4 DECLG 5 PC SG TUBE PLUGGING PRESSURE CORE BOTTOM () TOP , ( *)
2500 5
m E 2988 ISOS I Ns L-IBUS
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Ej 2 TIME ISCCI FIGURE 1 CORE PRESSURE - DECLG (CD=0.4)
NOTE: Asterisks (*) do not represent a separate curve, but provide a tracer to identify the curve associated with the peak (highest PCT) node. Where peak
, and burst node curves coincide, only one curve (with asterisk tracer will #
be seen.
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4 CD=0.4 DECLG CLAD AVG. TEMP. HOT ROD BURST. 7.25 FT( l PEAK 7.50 FT(*)
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TIME 15ECI FIGURE 2 PEAK CLAD TEMPERATURE - DECLG (CD=0.4)
NOTE: Asterisks (*) do.not represent a separate curve, but provide a tracer to i identify the curve associated with the peak (h1 hest PCT) node. Where peak
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CD=0.4 DECLG 5 PC SG TUBE PLUGGING I Z-FLOWRATE CORE BOTTOM () TOP , . ( *)
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NOTE: Asterisks (*) do not represent a separate curve, but provide a tracer to
} identify the curve associated with the z-flowrate at the top of the core.
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PUMPED ECCS FLOW FIGURE 4 PUMPED ECCS FLOW (REFLOOD) - DECLG (CD=0.4)
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