ML17341B045
| ML17341B045 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 04/15/1982 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17341B044 | List: |
| References | |
| NUDOCS 8204260202 | |
| Download: ML17341B045 (98) | |
Text
reac'Aj'vity inser~ upon ejection greater than O~ k/k at rated power. Inoper ab rod worth shall'e determined ~nin 4 weeks.
- b. A control rod shall be considered inoperable if (a) the rod cannot be moved by CRDM, or (b) the rod is misaligned from its bank by more than 15 inches, or (c) the rod drop time is not met.
- c. If a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate for the withdrawn worth of the inoperable rod.
- 5. CONTROL 'ROD POSITION INDICATION If either the power range channel deviation alarm or the rod deviation monitor alarm is "not operable, rod positions shall be logged once per shift and after a 1'oad change greater than 105 of rated, power. If both alarms are inoperable for two hours or more, the nuclear overpower trip shall be reset to 93$ of rated power.
- 6. - POWER DISTRIBUTION LIMITS
- a. Hot channel factors:
(1) F~ Limi t The hot channel factors (defined in Bases) must meet the following limits at all times except during low power physics tests:
Fg (Z) (~Fg~L/P) x K(Z), for P > 0.5 F~ (Z) < (2 x fF~]L) x K(Z), for P < 0.5 F < 1.55 f1.0 + 0.2 (1 - P)$
>H Where P is the fraction of rated power at which the core is operating; K(Z) is the function given in Figure 3.2-3 or Figure 3.2-3a; Z is the core height location of F. [F~]L and K(Z) are dependent on the steam generator tube plug)ing level as follows:
Plugging level [F~]L Figure Number for 'K(Z)
>5 g, and < 285 2.125 3.2-3
< 5$ 2. 30 3. 2-3a (2) Augmented Survei l lance (MIDS)
If $ F~], as predicted by approved physics calculations, exceeds
[Fq]L t3en the power will be limited to a turnon power fraction, PT, equal to the ratio of $ F]L divided by fF~] , or, for operation at power levels ab6ve .P7, augmented s(rveillance of hot channel factors shall be implemented, except in Base Load
- 3. 2-3. 4/15/82
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HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE (for ( 5% steam. generator tube plugging and F =2.30)
(10. 8, . 939)
(6.0,1,.000)
(12.0, .43 10 CORE HEIGHT'FT. )
FIG. 3.2-3a 4/15/82
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Sa.ety Analysis for Turkey Point Unit 3, and 4 Loss-of-Coclant-Accident The . s<-of-c="l=-nt ac'c.dent ,LCC~) has b en rea".,alyzed for Turkey Point 3 ~'" ~ -"-~-~a~~": ..;o"'el <4F steam ~e..e."ators and 5 percent steam gal:ei a o u =
p J4ging inis a i 'a~is iou IG a Iso Le appl cab i 2 to TU4ey Point Unit 4 (FLA) provided that it also has replaced model 44F steam generator. The following report amends ihe section of ihe Safety Analysis Report on major reactcr coo'.ant system pipe ruptures (Chapter 14.3.2).
Th analysis presented here is in accordance with the reouirements of 10CFR50.46, Appendix K provided in reference l.
A description of the Westinghouse Emergency Core Cooling System (ECCS)
Evaluation Model used for this analysis is presented in l!CAP-8839 (ref-erence 2).. The individual comouter codes which comprise the Westing-house ECC Evaluation Model are described in detail in references 3, 4, 5, and 6. The results of several sensitivity studies are reported in reference 8. These results are for conditions which are not limiting in nature-and hence are reported on a generic, basis.
Since the initial development of the Appendix K ECCS Evaluation Model, several model changes were'ade, submitted to the NRC for review and approved for use in design LOCA analyses. These modifications are specified in references 7, 9, 10, 11, 12, 13, and 14.
The LOCA analysis presented in this report utilized the: 1981 version of the evaluation model which is the model currently used and accepted for plant licensing calculations. The modifications which comprise the 1981 evaluation model are delineated in reference 12.
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Resul ts The analysis of the loss-of-coolant accident is performed at 102 percent of the licensed core power rating. The peak linear power and total core power used in the analysis are given in Table 2. Since there .is margin between the value of peak linear power density used in this analysis and the value of the peak linear power density expected during plant operation, the peak clad temperature calculated in this analysis is greater than the maximum clad temperature expected to exist.
Table 1'resents the occurrence time for various events throughout the accident transient.
Table 2 presents selected input values and results from the hot fuel rod thermal transient calculation. For these results, the hot spot is defined as the location of maximum peak clad temperatures. That 1'ocation is speci-fied in Table 2 for each break analyzed. The location is indicated in feet which is the elevation above the bottom of the active fuel stack.
Table 3 present's a summary of the various containment systems parameters and structural parameters which were used as, input to the COCO computer code[6] used in this analysis.
Tables 4 and 5 present reflood mass and energy releases to the containment, and the broken loop accumulator mass .and energy release to the containment, respectively.
Figures 1 through 17 present the transients for the principle parameters for the break sizes analyzed. The fol-lowing items are noted:
Figures lA 38: Quality, mass velocity and clad heat transfer coef-ficient 'for the hotspot and burst locations Figures 4A - 6B: Core pressure, break flow, and core pressure drop-The break flow is the sum of the flowrates from both ends of the guillotine break. The core pressure drop is taken as the pressure just before the core inlet to the pressure just beyond the core outlet Figures 7A - 9B: Clad temperature, fluid temperature and"core flow.
The clad and fluid 'temperatures are for the hot spot and burst location.
Figures 10A - llB: Downcomer and core water level during reflood, and flooding rate Figures 12A - 13B: Emergency core cooling system flowrates, for both accumulator and pumped safety injection Figures 14A - 15B: Containment pressure and core power transient Figures 16, 17: Break energy release during blowdown and the con-tainment wall condensing heat transfer coefficient for the worst break 1635Q:1
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(4)
Conc lusions Thermal Analysis For break up to and including the double-ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFR50.46 That is:
- 1. The calculated peak. clad temperature does not exceed 2200 F based on a total core peaking factor of 2.30.
- 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
- 3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The, cladding oxidation limits. of 17 percent are not exceeded during or after quenching.
The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity r emai ning i n the c ore.
1635Q: 1
ii (5)
References
- 1. "Acceptance Criteria for Emergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors"', 10CFR50.46 and Appendix K of 10CFR50.46. Federal Register, Volume 39, Number 2, January 4, 1974.
- 2. Bordelon, F. N., Massie, H. M., and Zordan, T. A., "Westinghouse ECCS Evaluation Model-Summary", WCAP-8339, July 1974.
- 3. Bordelon, F. N., et al., "SATAN-VI Program: Comprehensive Space-Time Oependent Analysis of Loss-of-Cool,ant", WCAP-8302 (Proprietary Version), MCAP-8306 (Non-Propietary Version), June 1974.
- 4. Bordelon, F. M., et al., "LOCTA-IV Program: Loss-of-Coolant Tran-sient Analysis", WCAP-8301 (Proprietary Version), WCAP-8305 (Non-Proprietary Version), June 1974.
- 5. Kelly, R. 0., et al., "Calculational Model for Core Refloodi ng After a L'oss-of-Coolant. Accident (WREFLOOD Code)". MCAP-8170 (Proprietary Version), MCAP-8171 (Non-Proprietary Version), June 1974.
- 6. Bordelon, F. N., and Murphy E. T., "Containment Pressure Analysis Code (COCO)", MCAP-8327 (Proprietary Version), MCAP-8326 (Non-Pro-prietary Version), June 1974.
- 7. Bordelon, F. M., et al., "The Westinghouse ECCS Evaluation Model:
Supplementary Information", MCAP-8471 (Proprietary Version),
WCAP-8472 (Non-Proprietary Version), January 1975.
- 8. Salvatori, R., "Mestinghouse ECCS - Plant Sensitivity Studies",
MCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Ver-sion), July 1974.
- 9. "Westinghouse ECCS Evaluation Model, October, 1975 Versions,"
MCAP-8622 (Proprietary Version), MCAP-8623 (Non-Proprietary Version), November 1975.
1635(}: 1
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- 10. Letter from C. Eicheldinger of Westinghouse Electric Corporation to D ~ 8 ~ Yassalo of the Nuclear Regulatory Commission, letter number NS-CE-924, January 23,, 1976.
- 11. Kelly, R. D., Thompson, C. M., et al., "Mestinghouse Emergency Core Cooling System Evaluation tlodel for Analyzing Large LOCA's During Operation With One Loop Out of Service for Plants Without Loop iso-lation Valves", WCAP-9166, February 1978.
- 12. Eicheldinger C., "Westinghouse ECCS Evaluation Model, 1981 Version",
WCAP-9220-P-A (Proprietary Version), MCAP-9221-A (Non-Proprietary Version), Revision 1, December 1981.
- 13. Letter from T. A. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, ~letter number NS-TMA-1981, November 1, 1978.
14- Letter from T. A. Anderson of Westinghouse Electric Corporation to R. L. Tedesco of the Nuclear Regulatory Commission, letter number NS-TtQ-2014, December ll, 1978.
1635():1
~i TABLE 1 LARGE BREAK TINE SEQUENCE OF EVENTS DECL Cp = DECL C = 0.6 p
(Sec) (Sec)
START 0.0 0.0 Rx Trip Signal 0.72 0.71 S. I. Signal 1.13 0.60 Acc. Injection 14.7 11,3 End of Blowdown 32.93 28.14 Bottom of Core Recovery 54.44 49.00 62.79 57.83 Acc. Empty 25.74 25.60 Pump Injection 32.93 28.14 End of Bypass
TABLE 2 LARGE BREAK DECLG C = 0.4 DECLG C = 0 6 D D Res ul ts Peak Clad Temp. F 2195 2053 Peak Clad Location Ft. 7,25 6.75 Local Zr/H20 Rxn (max)$ 6.07 3,968 Local Zr/H20 Location Ft. 7.25 7.5 Total Zr/H20 Rxn X < 0.3 0,3 Hot Rod Burst Time sec 45.2 45.0 Hot Rod Burst Location Ft. 5.75 6.0 Ca 1 cul a ti on 2200 NSSS Power that 102K of Peak Linear Power kw/ft 102% of 13.07 2.30 Peaking Factor (At License Rating)
Accumulator Water Yolume (I'~<<~ -'~<<'~~~~ 875 ft Fuel. region + cycle analyzed Cycl e Region uHZTI3 Al 1 UNIT ~4) ( If applicable) Al 1
<Qi TABLE 3 (Page 1 of 3)
CONTAINMENT OATA (DRY CONTAINMENT)
Net Free Volume 1.55 x 106 Ft3 Initial Conditions Pressure 14-7 psia Temperature 90.0 'F RMST Temperature 39.0 F Service Mater Temperature 63.0 F Spray System Number of Pumps Operating 2 Runout Flow Rate .l450 gpm Actuation Time sec.
Safeguards Fan Coolers Number of .Fan Coolers Operating Fastest Post Accident Initiation of Fan Coolers 26 secs.
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TABLE 3 (Page 2 of 3)
STRUCTURAL HEAT SINK DATA
'Materi al Thickness '( in) Area (Ft )
Paint 0.006996 Carbon Steel 0. 2898 87335.8 Carbon Steel 0.006996 1000086.0 Paint 0.006996 Carbon Steel 0.4896 35660.11 Carbon Steel 0.4896 12367. 5 Paint 0.006996 Carbon Steel 0.2898 Concrete 24.0 50430.0 Carbon Steel 0.2898 Concrete 24.0 16810.0 Paint 0.006996 Carbon Steel 1.56 4622.69 Carbon Steel 1.56 1540. 89 Paint 0.006996 Carbon Steel 5.496 1277.87 Carbon Steel 5.496 425.93 Paint 0.006996 Carbon Steel 2.748 951.525
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TABLE 3 (Page 3 of 3)
STRUCTURAL HEAT SINK DATA Wall Material Thickness (in) Area (Ft )
12 Carbon Steel 2.748 317.175 13 Paint 0.006996 Carbon Steel '.03 23550.0 Paint 0.006996 Carbon Steel 0.063 80368.5 15 Paint 0.006996 Carbon Steel 0.10 42278.25 16 Carbon Steel . 0.2898. 17190.0 17 Stainless Steel 0.032 113253.4 18 Stainless Steel 2.1264 3704.0 19 Stainless Steel 0.1398 Concrete 24.0 14392.0 20 Concrete 24.0 59132.0
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TABLE 4 REFLOOD MASS AND ENERGY RELEASES:
DECLG CD- 0.4 0
Time ('sec) "Total ( ibm/sec) Total (Btu/sec) 54.347 55.669 0.028 35.68 62.891 19.48 25135 74.291 32.84 42383 92.191 45.47 57824 111.591 62.32 78159 130.791 ~
75.63 94204 149.791 87.12 108032 186.891 285.03 161152 226.791 298.71 152313 271.091 312.18 141932 321.491 328.83 130016 382.191 342.69 115551 eoI n ~
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TABLE 4B REFLOOD tlASS .AHD ~ihERGY RELEASES DECLG CD = 0.6 Time (sec) iOTAL (ibm/sec) TOTAL (.BTU/Sec 49,.001 0 49,826 '0. 03 35.2 57.410 33.17 42713 68.529 34. 49 44423 85.929 50.14 63315 104.929 65.35 81692 124.029 77.64 96478 88.47 109506 142.929'79.929 274.99 162328 219. 329 291.11 154284'43981 263.229 304.64 313.329 321.21 132490 373.829,' 341.73 1'17812
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Table 5 II DECLG CD=0.4 Broken Loop Injection Spill During Blowdown T IHE HASS ENERGY ENTHALPY 0.000 3083. 402 183832.451 59.620
- 1. 010 2775.985 165504.238 '59.620'9.620 2.010 2549.479 151999.952
- 3. 010 2371. 235 141373.054 59.'620
'4.010 2225'.996 1327'13.868 59.620 5.010 2104.580 125475.082 59.620
- 6. 010 2000.289 119257.219 59.620 7.010 1909.287 113831.700 59.620
- 8. 010 1828.531 109016.989 59.620
- 9. 010 1756.087 10I697.907 59.620
- 10. 010 1690.537 100789.842 59.620
- 11. 010 1630.928 97235.939 59.620 12.010 1576.546 93993.651 59.620 13.010 1526.743 91024.431 59.620 14.010 1480.928 88292.925 59.620 1 5.010 1438.640 85771. 693 59. 620 16.010 1399.477 83436.827 59.620 17.010 1363.118 81269. 116 59.620
- 18. 010 1329.230 79248.708 59.620
- 19. 010 1297.278 77343.735 59.620 20.010 1267. 106 75544.886 59.620
- 21. 010 1238.699 73851. 229 59.620
- 22. 010 1212.124 72266.830 59.620
- 23. 010 1187.431 ?0794.611 59.620
- 24. 010 1164.479 69426.222 59.620 25.010 1142.940 68142.087 59.620
- 26. 010 1122.957 66950.723 59.620
- 27. 1104.421 65845.604 59.620 1086.744 64 791.649 59.620 010'8.010
- 29. 010 1070.187 63804.549 59.620
- 30. 010 1054.268 62855.469 59.620 31.010 1039.041 6194?.622 59.620
- 32. 010 1024.474 61079. 153 59.620
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Tabi e 5 8 DECLG CD=0.6 Broken Loop Inje Cion Spii) Ourina Biolvciolsn T lP P '4o J} 4 l. ~ I(CRI V TPA 0'M M~ ~ ~ ~ 8 'a II ~
2544.223 "0 'o~) ol i
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- 4. 010 2217. ".": ~ aL J ~ ~
S.O10 2094.SQ2 124874. 218 59.<23 6.010 'l988.200 118536.465 59.620
- 7. 010 1895.206 1"i2992.2Q9 0> ~ 6m.V 8.010 1812.9O7 108085.507 c,o
- 9. 010 I7o9 0~3 4' s V
'I 0. 010 1673
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'02 99756 '74 594 620
~1.010 16' <47 961.9.686 59.':0 12.010 1S59 682 929o8.245 5).620
- 13. 010 1~ .010 1466.596 a7438.'469 1 5.010 1425.293 84975.953 59.620 16.010 1386.597 82668.919 59.620
- 17. 010 1350.,397 80510.683 59.620
- 18. 010 131?. 382 78542.306 59.620
- 19. 010 1286.947 7.6222 7b>> 59.620'9.620 20 010 1258.891 75055,089>>
- 21. 019 1233.052 73514.588 59.620
- 22. 010 1208,918 72075.697 S9.520
- 23. 010 1186.570 70743 '21 59.520
- 24. 010 1165.640 69495,43Q 59.620
- 25. 010 1145.727 68308.260 59.620
- 26. 'I126.662 67171. 562 59.620 010'7.010 1108.630 66096.497 59.620
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IIIII':sll FPL 198! ECCS MOOEL O.l DECL'0=2.30 REPLACEO, STEAM GENERATOR 600.00 HEAT TRANS ~ COEFFICIEHT BURST ~ 5 ~ 75 FT( ) PEAK ~ 1 ~ 25 FTl ~ I 5on.oo ioo.oo 300.00 200.00 Ill'.sip 60.000 50.000 cD ceo 000
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FI'L ISB I ECCS HOOEL 0.6 DECLG F0=2. 30 REPLACEO STEAH CENERATOR 600.00 HE AT TRANS COE ffIC IE NT BURST ~ 6 ~ 00 FT( ) RE AK ~ 6 75 FT< ~ )
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TURKEY POINT(FPL) 0.6 OE(LC l98l MOOEL TO=2 ~ 30 REPLACEMEHT STEAM CEHERATORS MITH 0 PER(EHT CUBE <'UCCIHC PRE SSURE (ORE BOTTOM I I TOP, ( ~ )
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STATE OF FLORIDA )
) ss ~
COMTY OF DADE )
Robert E. Uhrig being first duly sworn, deposes and says:
That he is Vice President of Florida Power Light Company, the herein; That he has executed the foregoing document; that the state-ments made in this said document .are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Robert E. Uhrig Subscribed and sworn to before me this
~5. dav of 4
'GTARY'.PvUBLIC, n and for the County of Dade, State cf Flori a State of Rorida at Large My COmmiSSiOn eXpireS, Notary Public, Exing~ October 30, 1983 Bonded thru Maynard Bonding Agency
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