ML17340A813
| ML17340A813 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/05/1981 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17340A812 | List: |
| References | |
| NUDOCS 8103110672 | |
| Download: ML17340A813 (129) | |
Text
REACTGR'ORE't't.lER(QL AHO H'V t~(L EC" SAFETY" i Libel""<'8, THRE.".LOPP(('t'rRAT t,0y,~-.t-~I (~~T..-~~~~~~~~~~~~~t~~~~~~(~~~~~~~~t~~~" t'~~~~~~~~~~~(f~~~~~~~~~~~~~~~~~~~660'~--~--1-~~~~t (~'t'-~~"~~~.~t....t~~t f-~~~s t.:--~:1".:~": ":~~~'~~~::(t~~~~~~~"-~~~~~650 f t~~~t t~~t t~t I t t~~~~~~~~~~Q6QO's tt~'r~~t~~t O-'-~<-(~t~-.,-~--...~~<...I...--C: 4P I t:-"3 3CA"ge s p~:~~."~~"'-"""~~~.t..~--C=---~--~~~t'~~~~~~P t't~~t~*'t~~t~~t~~~~t~~~630~t~~~~t~~~~~~~~~<~~~~~~~~~~~~~~t-610 j~~t-~t 6DO 590 1TLg,.ps>a~~~~~~~~~~~~~.(.....:1 (.",.-,"........
"".t."."<~~~~t(l:-':-.':No(a
~ese curses are a'ppj icabie'wi(h:
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l;Stan<a generatpr tube piugging glq<ter~e<te:
- .-:(.::!::
~(:28 peiceri(::
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REACTOR COOLANT TEMPERATURE Overtemperature Q T<4T[K1-0.0107 (T-574)+0.000453 (P-2235)-f (Aq)l AT~Indicated AT at rated power, F 0 Average temperature, F Pressurizer pressure, psig f Qq)=a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers;with gains to be selected based on measured instrument response during startup tests such that: F<<(qt-qb)within+10 percent and-14 percent where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt+qb is total core power in percent of rated po~er, f (8,q)~0.For each percent that the magnitude of (qt-qb)exceeds+10 percerit, the Delta-T trip setpoint shall be automatically reduced by 3.5 percent of its value at interim power.0 For each percent that the magnitude of (qt-qb)exceeds-14 percent, the Delta-T trip setpoint shall be automatically reduced by 2 percent of its value at interim power.K1 (Three Loop Operation) 1-095*(Two Loop Operation)
~0.88+K1 1.095 for steam generator tube plugging<28 percent-2~3-2 3/5/81 I
Overpower/
T~To 1.11+-Kl dT K2 (T T)f (dt To~Indicat ed Tat rat ed power,F T~Average temperature, F T'Indicated average temperature at nominal conditions and rated power, F Kl~0 for decreasing average temperature; 0.2 sec./F for increasing average temperature K2=0.00068+for T equal to or more than T;0 for T less than T'ate of change of temperature, F/sec dT dt f (dq)=As defined above.Pressurizer
-Low Pressurizer pressure-equal to or greater than 1835 psig High Pressurizer pressure-equal to or less than 2385 psig.High Presssurizer water level-equal to or less than 92%of full scale-Reactor Coolant Flow Low reactor coolant flow-equal to or greater than 90%of normal indicated flow.Low reactor coolant pump motor frequency equal to or greater than 56.1 Hz.Undervoltage on reactor coolant pump motor bus-equal to or greater than 60%of normal voltage.Steam Generators Low-low steam generator water level-equal to or greater than 15%of narrow range instrument scale.+This factor is 1.11 for steam generator tube plugging<15%.This factor is 1.10 for steam generator tube plugging>15%and<19%.This factor is 1-08 for steam generator tube plugging>19%and<28%."This factor is 0.00106 for steam generator tube plugging>19%and<28%.2'3 3 3/5/8,1 6~DNB PAKQKTERS The following DNB related parameter limits shall be maintained during power operation:
a.Reactor Coolant System Tavg<578.2'F b.Pressurizer Pressure>2220 psia*c.Reactor Coolant Flow>268,500 gpm+With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce termal power to less than 5X of rated thermal power using normal shutdown procedures.
Compliance with a.and b.is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.Compliance with c.is demon'strated by verifying that the parameter is within its limits after each refueling cycle."Limit not applicable during either a THERMAL POWER ramp increase in excess of (5/)RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of (10/)RATED THERMAL POWER.'-.+Reactor Coolant Flow>268,500 gpm for steam generator pube plugging<15X.Reactor Coolant Flow>263,130 gpm for steam generator tube plugging>15/and<19X.Reactor Coolant Flow>255,075 gpm for steam generator tube plugging>19/and<28/.3.1-7 3 j5/.81 i
reactivity insertion upon injection greater than 0.3 b k/k at rated power.Inoperable rod worth shall be determined within 4 weeks.b.A control rod shall be considered inoperable if (1)the rod cannot be moved by the CRDM, or (2)the rod is misaligned from its bank by more than 15 inches, or (3)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 are not operable, rod positions shall be logged once,per shift and after a load change greater than 10%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 DISTRBUTION LIMITS a.Hot channel factors: With steam generator tube plugging<28%, the hot channel factors (defined in the basis)must meet the following limits at all times except during low power physics tests: F (Z)<(2.125/P)x K(Z), for P>.5 Fq (Z)<(4.25)x K(Z),'or P<.5 F H<1.55[l.+0.2 (1-P)]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;Z is the core height location of F.Xf F , as predicted by approved physics calculations, exceeds 2.123 the power will be limited to the rated power multiplied by the ratio of 2.125 divided by the predicted F , or augmented surveillance of hot channel factors shall be implemented.
3~2 3 3/5/81
HOT CHANNEL FACTOR NORt!AL IZED OPERATING ENVELOPE (for (28ll steam aenerator tube pluaaina and F=2.12")(6.0, 1.000)(11.2,.8<9)(12.0, A70)10 12 CORE HEIGHT (FT.)FIG.3.2-3 3/5/81 i
ATTACHMENT A LOCA ANALYSIS I
An overview of the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model which was developed in accordance with tge requirements of 10 CFR'Part 50, Appendix'K
'~), Xs presented'in WCAP-8339'~The individual C2p computer codes which comprise the Westinghouse ECCS Evaluation Model are described in detail in separate reports (all dated June, 1974)in references 3, 4, 5, and 6-Since that initial development of the Appendix K ECCS Evaluation Model, several model changes were made, submitted to the NRC for review and approved for use in design LOCA analyses.The"October, 1975" version of the Westinghouse ECCS Evaluation Model incorporates modifications specified in referenes 7, 8 and 9.Additional modifications delineated in references 10, 11, 1 2 and 1 3 update the model to the"February, 1978" version which is the model currently used and accepted for plant licensing calculations.
The LOCA analysis presented in this report was performed with an evaluation model which includes improvements made subsequent to the submittal of the"February, 1978" version model changes.These model changes include use of",,UHI.Sof,tware.Technology," ,and.;addresses, the problem of interaction, between the accumulator and pumped safety injection flow previously described in reference 14 and 15.There are, however, several differences between the Reference 14 method and that used in this analysis due to its application to a 3 loop plant-First, the split downcomer (elements 11-14 and 47-50)is divided into two azimuthal sections equal to 2/3 of the downcomer volume attached to the two intact loops and 1/3 of the downcomer volume attached to the broken loop.This modelling method gives a more realistic downcomer flow behavior during the transient than splitting the volumes equally.Second, the intact loop was not split at the pump junction into two legs as described in Reference 1 4-Instead, by attaching the two intact loops into the same downcomer region, the standard 3-loop plant loop nodalization can be retained.The nodalization used for the SATAN-VI calculation has been modified because of the differences presented above.The modified nodalization is shown in Figure 1 9-In addition, Turkey Point Units 3 and 4 have the safety injection lines connected to the accumulator surge line between the check valves-Due to the'agnitude of the accumul'ato'r'"'1'nje'ction'f1'ow".and
'thydraulic resistance of the check valve and piping from the accumulator line between the check valves to the cold leg, the pressure in the accumulator line will be significantly higher than the RCS cold leg pressure when the accumulutors are injecting.
Thus, the interaction between accumulator flow and the safety injection flow increases the injection pressure above RCS pressure.Current Westinghouse Evaluation Models assume that the safety injection pumps inject directly to the RCS presssure and not the accumulator line pressure.When the higher injection section pressure is accounted for, less safety injection delivery occurs and the calculated peak clad temperature will increase.Reference 1 5 presents details of the model changes required to allow safety injection into the accumulator line.This analysis includes revised component volume and heat tansfer area calculations based on methods developed for incorporation into an automated analysis~input processor.
Credit was also taken for revised accumulator line resistance including a general 6%%d reduction, and for the accumulator line volume.This analysis includes the removal of the 65'F conservatism on the initial steady state fuel pellet temperature.
This input change is discusssed
in WCAP-8720,"Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations", and approved by the NRC in reference 16.The modifications, in essence,.update, the"February, 1978" version of.the.SATAN-VI,code to include analytical techniques currently approved for use in the LOCA analyses of Westinghouse plants equipped with an Upper Head Injection (UHI)system.The modifications described in this report are, in the judgment of Westinghouse, in conformance with the requirements of 10CFR Part 50, Appendix K.As such, an evaluation model which includes those modifications is suitable for determining a core peaking factor limit which demonstrates compliance with the acceptance criteria set forth in 10CFR50.46.
The Loss of Coolant Accident for Turkey Point Units 3 and 4 has been reanalyzed with the above model modifications.
RESULTS Table 1 presents the occurrence time for various events throughout the loss of coolant accident transient.
The peak linear power and total core power used in the analysis are given in Table 2.Table 2 also presents selected input values and results from the hot fuel rod thermal transient calculation.
For these results, the hot spot location is specified for each break analyzed.The location is indicated in feet, which is elevation above the bottom of the active fuel stack.t Table 3 presents a summary of the various containment systems parameters and structural parameters which were used as input to the COCO computer code (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.
Quality, mass velocity and clad heat transfer coefficient for the clad location exhibiting the maximum temperature (hot spot)and for the section of clad that bursts, if applicable.
Figures 1 through 17 present the transients for the principal parameters for the break sizes analyzed.The following items are noted: Figures 1-3: Figures 4-6: 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)ust before the core inlet to the pressure gust beyond the core outlet.Figures 7-9: Clad temperature, fluid temperature and core flow.The clad and fluid temperature's are for the hot spot and burst locations.
l Figures 12-13: Emergency core cooling system flowrates, for both"accumulator
<<and pumped saf ety-.injection.
Figures 14-15: Containment pressure and core power transients.
Figures 16-17: Break energy release during blowdown and the containment wall condensing heat transfer coefficient for the worst break.Figure 18: K(z)versus core height for Fq 2.125.Figure 1 9: SATAN Model for Turkey Point CONCLUSIONS
-THERMAL ANALYSIS For breaks up to and including the double ended severance of a reactor coolantpipe, 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 2.25.2-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 t reactor 3~The clad temperat'ure transient is terminated at a time when the core geometry is still amenable to cooling.The local cladding oxidation limit of 17K is not exceeded during or after quenching~4.The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.5.As shown in.Appendix A,"it's e'stimated'that a"reduction in'F~o'f 0.].25 is required to maintain the 2200'F clad temperature as a result of the fuel rod models in NUREG&630.
This leaves a net F~of 2.125.REFERENCES 1."Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", 1 OCFR50.46 and Appendix K of 20CFR50.46-Federal Register, Volume 39, Number 3, January 4, 1974.2.Bordelon, F.M., Massie, H.W., and Zordan, T.A.,"Westinghouse ECCS Evaluation Model-Summary", WCAP-8339, July, 1974.3.Bordelon, F.M., et al.,"SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant", WCAP-8302 (Proprietary Version), WCAP-8306-(Non-Proprietary Version), June 1974-4.Bordelon, F.M-, et al-,"LOCTA-IV Program: Loss-of-Coolant Transient Analysis", WCAP-8301 (Proprietary Version), WCAP-8305 (Non-Proprietary Version), June 1974-I 5.Kelly, R.D.et al."Calculational Model for Core Ref looding after a~~~~~~~Loss-of-Coolant Accident (WREFLOOD Code)".WCAP-8170 (Proprietary Version),.WCAP-8171 (Non-,Proprietary Version), June,1974.
6.Bordelon, F.M., and Murphy E.T.,"Containment Pressure Analysis Code (COCO)", WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), June 1974.7.Bordelon, F.M.et al."The Westinghouse ECCS Evaluation Model: Supplementary Information".
WCAP-8471 (Proprietary Version), WCAP-8472 (Non-Proprietary Version), January 1975.8."Westinghouse ECCS Evaluation Model, October, 1975 Versions", WCAP-8622 (Proprietary Version), WCAP-8623 (non-Proprietary Vexsion), November, 1975.9.Letter from C.Eicheldinger of Westinghouse Electric Corporation to D.B.Vassalo of, the NuclearRegulatory Commission, letter number NS-,CE-924, January 23, 1976.10.Kelly, R, D.Thompson, C.M., et al."Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation with One Loop Out.of Service for Plants without Loop Isolation Valves", WCAP-9166, February, 1978.t 11.Eicheldinger, C."Westinghouse ECCS Evaluation Model, February 1978 Version", WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Non-Proprietary Version), February, 1978.12.Letter from T.M.Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-1981, Nov.1, 1978.13.Letter from T.M.Anderson of Westinghouse Electx'ic Corporation to Tedesco of the Nuclear Regulatory Commissin, letter number NS-TMA-2014, Dec.11;1978.14.Letter from T.M.Anderson of Westinghouse Electric Company to J.M.Miller, NRC, NS-TMA-2311, September 15, 1980.15.Letter from T.M.Anderson of Westinghouse Electric Company to D.F.Ross, NRC, NS-TMA-2354, December 22, 1980.16.Letter from J.F.Stol, NRC to T.M.Anderson, Westinghouse, dated March 27, 1980.
I 0 TABLE 1 LARGE BREAK-UHI TECHNOLOGY HOOEL TINE SEQUENCE OF EVENTS OECL)D--0.4}o~ci)D o.s)(Sec)(Sec)START I'RX'-Trip Setpoint'Reached 0.0 0.69 0.0 0.68 S.I.Setpoint Reached 0.78 0.60 Acc.Injection 14.6 11.2 End of Slowdown 30.31 25.07 Bottom of Core Recovery 50.213 44.867 Acc.Empty 61.35 56.731 Pump In'ject'i'on
'"25.78 25.60 End of Bypass 30.31 25.07 416A/0033A TABLE 2 LARGE BREAK-UHI TECHNOLOGY NOOEL OECL CO--0.4 OECL CO=0.6 Results Peak Clad Temp.'F Peak Clad Location Ft.Local Zr/H20 Rxn(max)percent Local Zr/H20 Location Ft.Total Zr/H20 Rxn percent Hot Rod Burst Time sec Hot Rod Burst Location Ft.2183 7.5 7.39 6.0<0.3 41.8 6.0 196Z.7.25 2.991 7.25<0.3 72.3 6.0 Calculation Core Power t4vt 102 percent of 2200 Peak linear Power kw/ft 102 percent of 12.77 Peaking Factor 2.25 Accumulator Water.Volume 875 ft/accumulator Fuel region+cycle analyzed Cycle Region UNIT 3 and 4 All Al 1 416A/0033A'
TABLE 3 LARGE BREAK CONTAINMENT DATA (DRY CONTAINMENT)
NET FREE VOLUME 1.55 x 10 Ft INITIAL CONDITIONS Pressure Temperature RHST Temperature Service Mater Temperature Outside Temperature 14.7 psia 90 F 39 F 63 F 39 F SPRAY SYSTEM Number of Pumps Operating Runout Flow Rate Actuation Time 2 1450 gpm 26 secs SAFEGUARDS'AN COOLERS Number of Fan Coolers Operating Fastest Post Accident Initiation of Fan Coolers 3 26 secs 416A/D033A TABLE 3 (continued)
CONTAINMENT OATA (DRY CONTAINMENT)
STRUCTIONAL HEAT SINKS THI CKNESS (INCH)AREA (FT2)Paint Carbon steel Carbon steel Paint Carbon steel Carbon steel Paint Carbon steel Concrete Carbon steel Concrete Paint Carbon steel Carbon steel Paint Carbon steel Carbon steel Paint Carbon steel Carbon steel Paint Carbon steel Paint Carbon steel Paint Carbon steel 0.006996 0.2898 0.006996'0.006996 0.4896 0.4896 0.006996 0.2898 24.0 0.2898 24.0 0.006996 1.56 1.56 0.006996 5.496 5.496 0.006996 2.748 2.748 0.006996 0.03*0.006996 0.063 0.006996 0.10 87335.8 1000086.0 35660.11 12367.5 50430.0 16810.0 4622.69 1540.89 1277.87 425.93 951.525 317.175 23550.0 80368.5 42278.25 416A/0033A
'
TABLE 3 (continued)
CONTAINMENT OATA (ORY CONTAINMENT)
STRUCTIONAL HEAT SINKS THICKNESS (INCH)AREA (FT2)Carbon steel Stainless steel Stainless steel Stainless steel Concrete Concrete 0.2898 0.032 2.1264"0..1398 24.0 24.0 17190.0 113253.4 3704.0 14392.0 59132.0 416A/0033A
TABLE 4 REFLOOD MASS ANO ENERGY RELEASES OECLG CO 0.4 UHI Technolo9>
Model TINE (Sec)A (Total)(LBM/Sec)NH (Total)(BTU/Sec)50.213 50.938 58.741 70.244 86.644 105.541 124.044 142.844 182.144 225.744 0.0 2.698-02 31.34 43.62 55.09 67.37 78.10 250.04 287.05 294.06 0.0 34.90 4.056+4 5.570+4 6.953+4 8.431+4 9.722+4 1.449+5 1.437+5 1.345+5 416A/0033A E
TABLE 4A REFLOOO MASS ANO ENERGY RELEASES OECLG"C>~0.6 UHI Technolo9>
Model TIME (Sec)M (Total)(LBM/Sec)MH (Total)(BTU/Sec)44.862 45.587 53.315 64.618, 80.418 98.318 116.718 135.118 173.818 216.718 0.0 0.0267 31.64 49.98 60.16 71.18 80.91 252.6 285.7 292.8 0.0 34.47 4.090+4 6.340+4 7.592+4 8.938+4 1.013+5 1.474+5 1.448+5 1.360+5 416A/0033A 1
TABLE 5 BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT CO=0.4 UHI Technology Model TIME~Sec)MASS FLOW RATE LBM/Sec)0.0 1.01 2.01 4;01 6.01 8.01 10.01 15.01 20.01 25.01 29.01 34.32 3195.049 2872.266 2635.418 2298.696 2064'529 1886.928 1744.883 1487.450 1313.054 1190.337 1126.647 1009.557 FOR ENERGY FLOW, MULTIPLY MASS FLOW BY 59.62 BTU/LBM.416A/0033A
TABLE 5A BROKEN LOOP ACCUMULATOR FLOW TO..CONTAINIENT CD-0.6 UHI Technology Model TIME~(Sec MASS FLOW RATE LBM/Sec 0.00 1.01 2.01 4.01 6.01 8.01 10.01 15.01 20.01 24.01 34.362 3195.049 2868.804 2628.985 2287.787 2048.893 1867.277 1723.220 1469.498 1304.673 1210.650 1010.478 fOR BIERGY FLOW, MULTIPLY MASS FLOW BY 59.62 BTU/LBM.416A/0033A 1
I.(000 1.2500 FPL/FLA DECLG CD*Oj)UUI 28PC SGlP LUSI IMIERACTIOH 2 PUHPS ISX 15 FUEL UPPER HEAD Al 1001 OUALIIY OF FI.UID BURSII 6.00 Ffl I PEAK~1.50 Flt~I LJ 4L lal 0 1.0000 O 0.1500 0.5000 0.2500 0.0 i 8 H)HQ~~~~~~~~a a a aaaaa Nm 3 3@33>1IHE (SEC)Pl 4 A ill t Cl Ill~O O Al O P1 O O O OOOO O O OOOO~~~~~~O 0 0 0 OOOO O O OOOOO In CO~Can~FIGURf 1 FLUID QUALITY DfCLG (CD=0.4)
I.iaoo 1.2500 FPL/FLA OECLC CO=0.6 UKI 28PC SCTP LHSI INTERACTIOH 2 PUMPS ISX 15 FUEL UPPER HEAO AT THOT OUALITY OF FLUID BURSTS 6.00 FT()PEAK~7.25 FT1~)>>laJ CL 1.0000 0.>500 I 0 0.5000 0.2500 0.0 0 0 0 0 l I RNNNImm)0 0 0 0 OOOO>>I I HIImlg Al i%V V1 EO>QN>>TIME (SEC)0 0 00000 8 8 0 00OO 0 000~~\~~~K.j 0 0 O OOOO r aA ao r asm>>O 0 0 O 0 O 0.0 N 0 0'a OOOO 0 0 OOOO~~~~O 0 Onnhn O O O OOOO V lh ED~CDIRT>>FIGURE 1A FLUID QUALITY DECLG (CP=0, Io)
50.000 FPL/FLA DECLC CO=0.<f UHI 2BPC SGTP LHSI IRTERACTIOH 2 PUHPS ISX I5 fUEL UPPER HE%0 AT TROT NASS VELOCITY BURST~6 00 f Tl)PEAK~1 50 f Tl~l CI 0 Cl Cl CI CI Iar CI 8 8 g 88g8 a o 0 ooao CI TIHE lSEC)a o 0 aaoao 8 o 8Do80 Cal e In all r man>>CI CI Iar CI 0 n O O QOQQO 0 0 OOOO~~~~~~Q 0 0 OOOQQ 0 0 DOQQD an lo IL cool>>FIGURE 2 HASS VELOCITY OECLG (Co 0 4)
50.000 LJ w lA I 0.0 L CA FPL/FLA OECLC CO=0.6 UHI 2BPC SGTP LHSI INTERACTION 2 PUHPS 15X15 FUEL UPPER HEAO AT THOT HASS VELOCITY BURST, 6.00 FTI)PEA)I 7.25 FTI~)-50.000 LJ 0~-100.00 C X-150.00-200.00$3 I)mmmm)Imm Na 3 mmI33B'J bl V lh EO a EOOP~TINE iM'C, O 0.0~~D 0 0, 0 Pl a 0 D OOOO 0 0 OOOO~~~~~~D 0 a D OOOO 0 0 0 OOOO~~g>aoacoch~" FIGURE 2A MASS VELOCITY OECLG (Cp 0.6) l (IIII':I(coo.oo 500.00%00.00 300.00 200.00 FPL/FLA DECLC CO=0.4.UHI 28PC 561P IHSI III1ERAC1101I 2 PUMPS 15K I5 FUEL UPPER HEAO Al 1001 HEA1'1RAHS.COEFFIC1EII1 BURSli 6.00 flI)PEAKED 7&flI~).,*.)II'.IIIII 60.000 50.000 lo.oao 30.000 ca.000 Alii 6.0000 5.0000 a.0000 3.oaoa 2.0000$.0000 Cl CI 8 1IHE ISECI FIGURE 3 llEAT TRANSFER COEFFICIENT DECLG (CD=0,4)I CI C)CI A CI V1 Cl CI CI CU
0 600.00 500.00<00.00 300.00 Ai 200.00 FPL/FLA DECLC CD=0.6 UHI 2BPC SCTP LHSI INTERACTION 2 PUMPS 15K 15 FUEL UPPER HEAD AT THOT HEAT TRANS.COEFFICIENT BURST~6.00 FT()PEAK~7.25 FTl~)~i0.000 v 30.000~2O.OOO A)))6.0000 5.0000 i.oooo 3.0000 2.0000$.0000 CI C)T IHE (SE C I FIGURE 3A HEAT TRANSFER COEFFICIENT DECLG (CD~0.6)
C~iN].D fPL/flA OEClC CO=9.a Ultl 2BPC SClt'HSl lttTEttAC110tt 15X l5 fuEI, UPPEIt HEAD.Al lttD1 PttESSURE': ltt gl.lt 1, at.i~PUDD.O V)Q 15DD.0 1000.0 500.00 0.0 f 1 ME 1 SE C l P FIGURE 4 PRESSURE OECLG (CD=0.4)CS C)CI Cl P CS C)CI~A
fPllflA DEClC CD=0.6 Util PSPC'SC1P lltSI IHtlAACIIDN t5<1$futl uPPEA NEAO At twot PAE SSUAE IN El It~to IIt.i;,rp V7 v7 QJ L Q I000.0:00.09 CI CI CI CI CI t I HE t SE C I CI CI~Vl CI E3 CI CI C1 CI FIGURE 4A PRESSURE DECLG (CD=0.6)
I~~l.~~65~l i.: q)}': i sIlg~.k~l..r tPl.>tlh 9(CLC CONG.W Nil C8P}j: SClP LN)l lNtiihtllON l'w lS tUE)" iiP ('plHt AO h1.IHol 08th(f(fg,::.'...'5.'.I t r'~%.0XR~I~~~e<'I i~'I I i" I)~~~,;~.'l,~.--,\~.~~~" I~)'!:.!'I}~,}.!."!~$~I~5(2)0'~~j*~l.5&II'l Ch a'~~,$;~}}Q llHt iS.t'ai~~fIGURE 5 BREAK FLO!3 DECLCi (Cp=0,4)~.j
I.OOE F05 U 1.50E+04'IJ v7 n 5.0OE.O<fPLIfLA OECLG CO=0.G Ulll 28PC SCTP LllSI IHTERACTlOll ISX I5 fUEL UPPER HEAO AT TllOT BREAK fLO'M 51()2.5OEI01 0.0-2.50EI01'5.00EI01-l.50E I01-I.OOE~05 C$I I HE I SE C I FIGURE 5R ORERK FL017 PECLG (Cp=0.6)a CI AJ CI CI C)V'I~IJ'
)I)0.00)0 So.oo'o~Q i-FPL/FLA OKCLG CO=0.i UHI 28PC SCIP LHSI INIERACIIDII ISX IS FIIEL UPPER HEAD A1 THOI.CORE PR,ORQP III EL SI).I6(.I R 25.000 0.0-25.000 50.000-.10.PPQ R I e A D 8 O CU llHf ISfCI CI ID C5 FIGURE 6 CORE PRESSURE PROP PECLG (Cp=0,4)I o,
10.000<PL/ELA OECLG CD*0.C UHl 2llPC SC1P LHSl lklEAACT10II lSM l5 fUEL UPPEII Nf AO Al'IHOl COAE PA.QIIOP ill EL St),foal,)50.000 U7 Q cc 25.000 CY EJ 0.0-25.000-50.000.-)0.000 CI Ce I!IIE lSEC)FIGURE 6A CORE PRESSURE DROP DECLG (CD 0~6}
2500.0 FPL/FLA DECLC CD*0, Ulll 28PC SCTP LIISI IIITERACTTOII 2 PUHPS ISX I5 FUEL UPPER READ AT TROT CLAD AVG.TEHP.IIOT ROD BURSTc 6.00 FTI I PEAK, 1.50 FTI~)cL C000.0 o o CL IS00.0 Ll o I000.0 500.00 0.0 o a a o TlHE ISECI 8 Q CI a.a Ilv FIGURE 7 PEAK CLAD TEl'1PERATUBE DECLG (C=0,4)D
2500.0 TPL/fLA DECLG CO=0.6 UIII 28PC SCTP LUST THTERACTTOH 2 PUMPS 15>IS TUEL UPPER HEAD Al THOl CLAO AVG~TEMP.HOT AOO BURST 6.00 flI I PEAK, 1.25 fTI~I 2000.0 o CD CC ISOO.O.Z E I cD I000.0 Ll 500.00 0.0 CD o TTME ISEC)8 8 FIGURE?A PEAK CLAD TEMPERATURE DECLG (CD=0.6)
2000.0 a7.I)50.0 FPL/FLA OECLC CO=0.)if URI 20PC SC1P L)ISI I)IIERACIIO)l 2 PUHPS 15x 15 FUEL uPPL'R NEAO AI 1))01 FLUIO 1EHPERA1URE BURSf o 6~00 Fll)PE AK~7 50 Ffl+)I 500.0 I250.0)000.0)50.00 500.00 250.00 0.0 C)8 8 1I HE IS E C l FIGURE 8 FLUID TEMPERATURE DECLG (CD 0 4)CI Cl~ll
2000.0 u.)750.0 FPL/FLA OECLC CO=0.6 UHI 28PC SCTP LHSI IHTERACTIOH 2 PUMPS ISX 15 FUEL UPPER HEAO AT THOT FLUIO TEHPERATURE BURSTS 6.00 FT1)PEAKED 7.25 FT1~)El o 1500.0 1250.0 a 1000.0 X)50.00 500.00 250.00 0.0 C)TIHE 1SEC)FIGURE 8A'LUID TEMPERATURE DECLG (Cp=0.6)O O
~..*....j. ~I r~g~FPLJFI,A P(5<(ffll0.4 Ulll 28PC SCOP LUSl Itl1EAACllON l5X t5 FUEL;:UPP)P.klEAP 'Al Toot"fLOMflATE ':)Q,,E'.:t)lI.1(,).1 I~2)00.)I Cl 1 I~".5 l~~I:I'I EE I'E.,)$l r'~~,~~~-.I:l El~',.~5,l f."I.E~I 0.0~,I E'E~~-2500.P-saoo.o~'1~-7000.0 I:~'Cl I I TIHE<SECI Cl Cl Cl CI IEI FIGURE 9 CORE FLOH RATE (TOP ANO BOTT+'1 j DECLG (Cp=0 I)
1000.0 J 5000.0 07 n<PL'ISOLA DECLC CO=0.6 UHl 2SPC SCOP LHSl l4lEAACf f04!ST 15/UEL UPPEA HEAD AE TH01 2-ELOMAAfE t4 EL ft l lt,)2SOO.O 0.0 8 G TTO/vl-2500.0-SOOO.O~-7000.0 ltHE (SECl.FIGURE 9A CORE FLOltl RATE (TOP AND BOTTOM)DECLG (CD=0,6)
20.<<0 I1.500 FPLIFLA OEcLG co=0.UUI 28Pc sclP Lllsl IHIERAclIOH a PUHPs I 5'X I5 FUE L UPPE R UE AU A I IUOI MAIE.R I.E V EL(FI)L OLVtJ(C.F.Li I-.l5.000 I2.500 l0.000 (g.iI': i S.oooo 2.5000 0.9 CI C)C5 8 C)CI TIHE ISECI FIGURE 10 DOHHCONER AHD CORE tlATER LEVEL DURIHG REFLOOD TRAHSIEHT DECLG (Cp=0')
20.000 11.500 FPL/FLA 0ECLC CO=0.6 UHI 28PC SGTP LHSI IHTEIIACTIOII 2 PUMPS Isx 15.FUEL UPPEA HEAD AT THOT MATER LEVEL(FT)Li<'-i<>C L'-"'=I 1 000 12.500 10.000 I\~1.5000 C 6:I-"-5.0000 2.5009 0.0 CI 8 S TIME ISECI FIGURE lOA DOHNCOHER AND CORE HATER LEVEL DURING REFLOOD TRANSIENT'ECLG (Cp=O.6) 't I I 2.0000 l./5GD FPL/FLA OECLG CO=0 S4-UHI 28PC SGTP LHSI INTERACTION 2 PUHPS I5X l5 FUEL UPPER HEAD AT THDT FLOOD RATEIIN/SEC) I 5000 l.2500.I.0000 0.7500 CI CI 0.5000 0.25'00 0.0 CI CI 8 S TIME lSECI 8 8 m CI CI CI CS III FIGURE 11 CORE INLET VELOCITY DURIHG REFLOOP TMHSIEWT PECLG (Cp=0.4) O I LJ'CC CL I Vl I Cl<x CL I CO A I ZO X O II O<~cA VOX'~~CC O C7~~o CI.~~~4 00 001 4 C)O X C9 4 0 II I-V O O LrJ Cg CD I Ltl LaJ 4 I 4J LrJ'O V cC 0 0 O C7 CI C).CI ()35/8l)31't0 00019 ED CI O O
0't , l$090.0'-)~'h EPL/fib 0(Cl.G:,$0=0.%U))l 28PC SG(P Ll)Sl l))lEAACTIO)).'-.i.;I.'I ",",.:,'SX LS FUEI,"PPPtf R PEh0 AT THOl;: II):.f)..43();5).)I)3<~), 6(+)h"~~~~I~-1l t000.tj'-~.~l~I I;I'000.$!'r IIOOO,O i,~.h~q I.~Jh l.,~l: f'~0 I~~h0 l".<"I.'i:.:,]". lh~..I[~h~'~I I I'j Iooo.)..I.'*,, I~~~~~~~I 1~'l~)~I~~r.I~t~I II t I'h~t I'~'.~0.0~,'A I~t'I~0 ,hl TAHE l)EC)Ci CS 0tl~~I I 0 I~C0 I'I,t~~C5:.,i'l~FIGURE qP gKUgULATOR INJECTION DURING OIOHDONN DECLG (CD=O.4)
5000.0 fPLIFLA DECLC CDIO.C UHl 2IPC SClf LHSI INTEAACll08 ISX l5 fUEL UPPER HEAD'l lHOl IH EL Cjt>.Oi.).i)led. Clii 4000.0 V7 Irl 3000.0 2000.0 l000.0 0.0 a a a lltiE iSECi FIGURE 12A RCCUHULATOR IHJECTIOH DURIHG OLOHOOW.DECLG (CD=0.6)
FIGURE 13 PUMPED ECCS FLO!t (REFLOOD)DECLG (CD'4..-'I a~a a~" a~a~~a~o~a a l a'I aa~'Ji, r Z 4'J=F.4 0'5 laa 4 X 4I~a a a i a I~I~I'a,'I~!" i a I 1/~I a I~[~I'I'I'I'a a a I'll i~~~~I Ca: j I'I,.I....;."-.,'j!'yi:.;I/'-:.-..::.; j"K II::I I.I'I I.'!<<~~[l: I-I: I l i I I 1'I i.I!I-.a I-'I~.~I~
at g ltt tS tit IO I III.I I I I I ll.ll I I It~Ill III I I I Cl t at II~.\~sara>>4>i~r I)IIII f~~a I*I il Cl m~a I I I I I I I I O Z m D A m r O m A.A A O Ch ll l O O Ol m I C)O 1~~~\*I I Ill ac~~s~::~I~I~I:I~~si~l st>Is I;..:~>>~ar l*>>.I I~~~t, I II l'I>>: l.: I.I>>I>>: I!I~I s 1 I I I I.II lss~~~a'll s~.t~i t~~tl.'I-I I'i"'s i la I'(I a;!'I I>>I;'iris~s'I~I.I>>'I s s stt'I: s~I.I I s I I ll I:I I il I lo I<.ltl IO I}IF.Cl I lilt t}.IFII t ttLUtlCt lt CS'tt tt~:It~III~<<'v<.I t'>>I'.i II)F~I I I I't<<'.I'I/I~~I I'I'~t<I~I Tl Cl Pl C)O Z Pl O I M Pl n II D O pl Vl Vl Kl Pl I~',I I F g.ICI I'Q~~~~~~t-'I<I<'}I~~-c'0
lu X 10 10 fl!C Cf:NTI!sll.llsl! I'll'V4 IILUI ffI.II f5&II I II'usss~~ss~!: I!ill!ps 5 3P I I%5!I~~~-~)I I~II I"~ii jf ,"~I I!'!il+~!I;':si.!I~~~~~I~I!j, I j I I I j.I I~~I i sj I~~\I I I l j::I J~:.'.,'I I I I~~~m f n O O m p 0 C)I-I n m II a O Xl~m tn m Is-j I I',s..if/I*~~'~~"I I'I".i~~~
I~I.oooa=too I<~t g~, t t.I~t'P:I)I t..FPI./flA gFCI.C cg~a,t u>>l 20Pc sCtP Iwsl l>>tcRActlaN ~::,,',..lsx 15 t'ural UPPER IIEPO.at t>>ol I.':."~".'~." OVfR;.': 52I)0.8000"~'tt~t'rttt I~0.6)0)l o.saon..~5 It ,q~~It~t'~~'II I r~15 I~~~~~---5'~I.II\..II)'.ti5'~~,I.':l'.~~>I I I 0.2000 I I.I'~0.0 I~e'~'~~C1 I tlI>r tsfci ED C)CI~o O I,I~,-CI5 C'I'I FIGuRE 1s CORE POWER TRnHsIEHT oECLG (CD=O.d)
1.0000=I+0 la FPL/FLA DECLG COCO.6 Ulll 28PC SGTP LIISl IIITERACTIOII ISX I5 FUEL UPPER IIEAO AT TIIOT POMER 52l I 0.8000 o.6ooo L~0 O.looo 0.2000 0.0 CI CI C3 CI V1 I.I 1 CI TTNE ISEC)CI CI Ill CI i'I CI CII CI CI CI llI FIGURE 15A CORE Ppl<ER TRANSIENT PECLG (Cp=0,6) 5,00E F01 3.0DE ql'CQ s I I I I I.t~J.~FPL/FLh PE(L/" CO=0;i Ulll 28PC SCTP LIISI lIITEAACTIOII TSX IS.FUEI, UPPERS IIEAP hT TIIOT QREAK" EIIEQQV: f.:;, 5%l I'.1.00E ill I~-l.OOEiOl 3~OOErOT, I 1 I P~.', I.-.~~F'g*~~~~'I J~t.-.l'~~~'I.5;00fr01 I I li I~Cl C)TTHE I'SfC)s FIGURE 16 BREAK ENERGY DECLG (t:>-P g)
J C C 5-OOE Col CLJ v7 3.00ECOT CQ FPLIFLA OECLC CO=0.6 UIII 28PC SCTP LIISI IHTEAACTIOII 15K I5 FUEL;UPPEII IIEAO AT TIIOT BREAK E IIE RG Y 50 l Cl.Cl CD CI Cl IA CD TINE ISE C I Cl CD CCl Cl CD Cl CCC Cl Cl Cl~Cl FIGURE 16<BREAK ENERGY PECLG (Cp" 0.6)
Iit X 10 lO Tlif I:I.'IITInII'.TEIY ~'~'I Fe I<lilt FCI.&f SliCIC l:0 IW1 I>>II I I I 0 I('I~~I~"I: I~:;I'i:~I.~-I~.I I'I~AJ t I I I'I~~~.I~~~" I'~~I.t::, t~~11'~I~I~'~~~'i..;I~~I I I.*..I I i I I l:I I I';~i.Tl Cl Pl n I n II C)Tl Kl TTl n ED Pl-i F I I Tl TFi Kl n C)TTl Tl Tj n g I I~II~~~~~t I Iit X Iii'fO 1'IIC C,CNfl(If< ff(t I'.i'".~~I I'F~Et I<I'UIFCI. <(CSSO<<;<(tttt'I(jf(I()t~~I)~H!C'f" (i i 1"'r I t , I-'Oi t I (: c F'i ti I I I I,:I: I("~~.I I rl Rl m~~~~.~~t~~I I I i I~~i t I I i I I I t-I~i tl~i~I~~t'I Tl I C)m n I n tD II CD n ED t i m I>~I m Vl ll n C3 m ll Tl m t t~i~~~~'t"'t'I~~I I t)
HOT CHAl'hlEL FACTOR t{ORl!AL IZED OPERATING EllYELOPE (7or<28~steam aenerator tube pluaaina and F=2.12)(5.0, 1.000)(11.2,.849)(12.0,.470)10 CORE HEIGHT (FT.)
('SrE;~t GEttE(t~rottz gQ'>N NO R PTP 3 5 4 (54 El ENENTS)ST%P*;GCMAMOtt:-./~3G 83 YESSB.t2 t ttEswftIKEtt .<K~6..23 3A 37'l l.I+7 I J I)8'>g f 9.lf 7'2..l3 lif.30~~29'l~I tq 2Q Ho.tt-t tttrAcr ILobvs z 4C<~ie~t5 FIG.19 gg 2?: 0tt0Kot lo".t',COllTA I tHIEH I'NODE I'~I I: I/I C
=APPENDIX A The Nuclear Regulatory Commission (NRC)issued a letter dated November 9, 1979 to operators of light water reactors regarding fuel rod models used in Loss of Coolant Accident (LOCA)ECCS evaluation models.That letter describes a meet-ing called by the NRC on November 1, 1979 to present draft report NUREG 0630,"Cladding Swelling and Rupture Models for LOCA Analysis." At the meeting, representatives of HSSS vendors and fuel suppliers were asked to show how plants licensed using their LOCA/ECCS evaluation model continued to conform to-10'CFR Part.50-46 'in**view-of the'new fuel rod.models. presented in draft HUREG 0630.Westinghouse=representatives presented infor'mation on the fuel rod models used in analyses for plants licensed with the Westinghouse ECCS eval-uation model and discussed the potential impact of fuel rod models used in analyses for plants licensed with the Aestinghouse ECCS evaluation model and discuss'ed the potential impact of fuel rod model changes on results of those analyses.That information was formally documented in letter HS-'MA-2147, dated November 2, 1979, and formed the basis for the Westinghouse conclusion that the information presented in draft NUREG 0630 did not constitute a safety problem for Westinghouse plants and that all plants conformed with NRC regula-tions.In the November 9, 1979 letter, the HRC requested that operators of light water reactors provide, within sixty (60)days, information which will':enable"'the 'staff.'to"determine,'in light'of'the'fuel rod model concerns, whether or not further action is necessary. As a result rf compiling information for letter HS-THA-2147, Westinghouse recognized,a potential discrepancy in the calculation of fuel rod burst for 416A/0033A
cases having clad heatup rates (prior to rupture)significantly lower than 25.:degrees F per;second. Thi;s i.ssue.was reported.to..the.NRC staff, by, telephone, on November 9, 1979, and although independent of the NRC fuel rod model concern, the combined effect.of this issue and the effect of the NRC fuel rod models had to be studied.Details of the work done on this issue were pr e-sented to the NRC on November 13, 1979 and documented in letter NS-Ti%-2163 dated November 16, 1979.That work included development of a procedure to determine the clad heatup rate prior to burst and a reevaluation of operating Westinghouse fuel rod burst model.As part of this reevaluation, the Westinghouse position on NUREG-0630 was reviewed and it was still concluded that the information presented in draft NUREG-0630 did not constitute a safety problem for plants licensed with the Westinghouse ECCS evaluation model.On Oecember 6, 1979, NRC and Westinghouse personnel discussed the information thus far presented. At the conclusion of that discussion, the NRC staff requested Westinghouse to provide further detail on the potential impact of modifications to each of the fuel rod models used in the LOCA analysis and to outline analytical model improvements in other parts of the analysis and the potential benefit associated with those improvements. This additional infor-mation was compiled from various LOCA analysis results and documented in'etter NS-TMA-2174 dated Oecember 7, 1979.Another meeting was held in 8ethesda on Oecember 20, 1979 where NRC and Westinghouse personnel established: 1)The currently accepted procedure for assessing the potential impact on LOCA analysis results of using the fuel rod models presented in draft"NURFG-0630 and 2)Acceptable benefits resulting from anlytical model improvements that would justify continued plant operation for the interim until differences between the fuel rod models of concern are resolved.416A/0033A 4 Part of the Westinghouse effort provided to assist in the resolution of these iLOCA fuel.rod model idi.fferences.is, documented in.l,etter NS-TMA-2175,,dated December 10, 1979, which contains Westinghouse comments on draft NUREG-0630. As stated in that letter, Westinghouse believes the current Westinghouse models to be conservative and to be in compliance with Appendix K.A.Evaluation of the potential impact of the fuel rod models presented in dr aft NUREG-0630 on the Loss of Coolant Accident (LOCA)analysis for Turkey Point Units 3 and 4 with 28 percent SGTP.This evaluation is based on the limiting break LOCA analysis identified as follows: BREAK TYPE-DOUBLE ENDED COLO LEG GUILLOTINE BREAK DISCHARGE COEFFICIENT CO=0.4 WESTINGHOUSE ECCS EVALUATION MODEL VERSION Revised February 1978 with Safety Injection Interaction Modifications and UHI software technology CORE PEAKING FACTOR 2.25 HOT ROD itAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD 2041'F-PCTB ELEVATION HOT ROD MAX'IMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD-2183 F=PCTN FLEVATION-7.5 Feet 0 CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION 1.93 Percent MAXIMUM CLAD STRAIN AT THIS El EVATION-9.04 Percent 416A/0033A
Maximum temperature for this non-burst node occurs when the core reflood rate is'GREATER than 1.0 inch per second and reflood heat transfer is based on the ELECHT'alculation. AVERAGE HOT ASSEMBLY ROD BURST ELEVATION-N/A Feet HOT ASSEMBLY BLOCKAGE CALCULATED -0.0 Percent-1.BURST NODE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in the peaking factor 1'imi't'(FQ)required'to maintain a.peak clad temper'atur e (PCT)of 2200 F and in terms of a change in PCT at a constant FQ.Since the clad-water reaction rate increases significantly at temperatures above 2200.F, individual effects (such as aPCT due to changes in several fuel rod models)indicated here may not accurately apply over large ranges, but a simultaneous change in FQ which causes the PCT to remain in the neighborhood of 2200.F justifies use of this evaluation procedure, From NS-TMA-2174: For the Burst Node of the clad: '0."01'aFQ+-150 F'BURST'NODE aPCT Use of the NRC burst model and the revised Westinghouse burst model could require an FQ reduction of 0.027 The maximum estimated impact of using the NRC strain model is a required FQ reduction of 0.03.416A/0033A
Therefore, the maximum penalty for the Hot Rod burst node is: hPCT1-'(0."027'+.03)(1'50 F/.Ol)=,855 F Margin to the 2200 F limit is: t)PCT2 2200.F PCTB 159 F The Fg reducion requir ed to maintain the 2200 F clad temperature limit is: aF0=(aPCTl-aPCT)(~)=0.0464 (but not less than zero).2.NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the core mid-plane during the core reflood phase of the LOCA transient. The potential impact on that maximum clad temperature of using the NRC fuel rod models can be estimated by examining two aspects of the'analyses. The first aspect is the change in pellet-clad gap conductance resulting from a di'fference in clad strain at'the non-burst maXimum clad temperature node elevation. Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculate'd Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase resulting from a change in strain (in the Hot Rod)is+20.F per percent decrease in strain at the maximum clad temperature locations. Since the clad strain calculated during the reactor coolant system blowdown phase of the accident is not changed by the 416A/D033A 4 use of HRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference between the"maximum clad strain" and the"clad strain at the end of RCS blowdown" indicated above.Therefore: 4PCT3=(~0<.)(NAX STRAiN-BLOWOOWN STRAIN)=(+I)(0.0904-0.0'l93)142.2 F'I The second aspect of the analysis that can increase PCT is the flow blockage calculated. Since the greatest value of blockage indicated by the HRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above)is raised to 75 percent and then applying an appropriate sensitivity formula shown in HS-T)MA-2174 .Therefore, aPCT4-1.25 F (50-PERCEHT CURREHT BLOCKAGE)+2.36 F (75-50)=1.25 (50-0.0)+2.36 (75-50)121.5 9 If PCTH occurs when the core ref lood rate is greater than 1.0 inch per second aPCT=0.For Turkey Point PCTH does occur when the core reflood rate is greater than 1.0 inch per second.The total potential PCT increase for the non-burst node is then aPCTS aPCT3+(4PCT4=142.2+0=142.2 416A/0033A
Margin to the 2200 F limit is aPCT6=2200 F=PCTN=2200-'183'17 F The FQ reduction required to maintain this 2200 F clad temperature limit is (from NS-TNA-2174) AFDN PCTS-APCT6~DF PCT.OlaFQ aFQN-0.125 but not less than zero.The peaking factor reduction required to maintain the 2200 F clad"temperature limit is therefore the-greater of aFQ8"and'aFQN, QPEHALTY 8.The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate revised Hodel FQ credit identified minus the aFQPE>ALTY,calculated in section (A)above (but not greater than zero).FD ADJUSTNENT =0.125 Final FQ Yalue=2.25-0.125 2.125 416A/0033A 4 ATTACHMENT B NON-LOCA ANALYSIS C INTRODUCTION The NSSS vendor has performed an evaluation of the impact of increasing the steam generator tube plugging Ievel to 28 percent and of increasing the peaking factor F~to 2.1 75+on the non-LOCA accident analyses presented in Chapter 14 of the Turkey Point Unit 3.and Unit 4 FSAR.The current Turkey Point safety analyses are valid for steam generator tube plugging levels up to 25%, with>95%of RCS thermal design flow, and a peaking factor of 1.93.CHANGE IN STEAM GENERATOR TUBE PLUGGING LEVEL Changes in the number of steam generator tubes plugged result in lower reactor coolant flow rates.Many of the non-LOCA accidents are strongly dependent on the flow tate available. Recent flow re~ting has shown that the previously assumed flow rate of 85025 gpm per loop, corresponding to 95%of RCS thermal design flow rate, conservatively bounds the predicted flow with 28%tube plugging.Therefore, effects due a change in flow rate need not be'ddresssed. Because the time duration of the non-LOCA transients" are very short, changes in pressure drops and time response due to tube plugging have'little'effect on'the'c'ourse of rtrans'ients'/acc'idents. 'Increasing the tub'e plugging level to 28%decreases the heat transfer area available in the steam generator. The impact of this, on the non-LOCA accident analysis also is not signif icant.CHANGES IN Fq LIMIT The impact of higher F~values, on the non-LOCA accident analyses presented in Chapter 14 of the FSAR was analyzed-The accident/transients considered were:-Uncontrolled RCCS Withdrawal f rom Subcritical Conditions.-Uncontrolled RCCS Withdrawal at Power-Malpositioning of a Part Length Rod'-Rod Clus'ter Control'A's'se'mb'ly '(RCCA')'Drop-Start-up of an Inactive Reactor Coolant Loop-Reduction in Feedwater Enthalpy Incident-Exessive Load Increase Incident-Loss of Reactor Coolant Flow (Locked Rotor Accident)-Loss of Reactor Coolant Flow (Flow Coastdown Accident)-Loss of External Electrical Load-Loss of Normal Feedwater-Loss of AC Power-Rupture of Steam Pipe e*'Proposed LOCA analysis requirement sets the limit on F at<2e125, as per-Attachment A.
t Rupture of Control Rod Mechanism Housing Chemical and Volume Control System Malf unction The effect of flow reductions and changes in the value of the peaking factor, F, on the above accidents are discussed in detail in a previous analysis (eference 1), which was performed for 25 percent steam generator tube plugging and 95X reactor coolant flow-As the flow rate with 28X steam generator tube plugging is not result in a flow rate less than 95%of design flow, changes in need not be addressed and previous discussions and conclusions (reference 1)are still valid.However, changing FQ from 1.93 to be considered. expected to the flow rate p r esented, to 2.175*needs From the accidents presented in Chapter 14 of the FSAR, only the following accidents are affected by changing the value of FQ'.'"LOSS"OF"REACTOR"COOLA'NT""FLOW " LOCKED'ROTOR ACCIDENT The current applicable analysis of this event is based on a hot spot heat transfer calculation which utilizes heat flux and fuel temperatures associated with an FQ of 2.55~An FQ limit of<2.1 75 results in a 14%reduction in (2)total energy input to the hot spot which will compensate for the reduction of coolant flow to 95X of thermal design flow of 85025 gpm per loop, due to 28X tube plugging.This will result in peak fuel and clad temperatures well below the safety limit.Hence a change of FQ f rom 1.93 to 2.175 has little effect on the safety limit margins and the conclusions of reference 1 are valid for 28X level tube plugging.RUPTURE OF CONTROL ROD DRIVE MECHANISM HOUSING The applicable current analysis was performed for an FQ of 2.32 at 100X flow rate, (reference 2)~From LOCA considerations FQ has to be below 2.175*.Therefore, changing the value of FQ from 1~93 to 2.175 still results in a.reduction in, initial,f uel temperature. which.,translates,,into a,reduction.in peak transient f uel temperatures compared to those in reference 2.Also according to reference 1 reducing the coolant flow to 95X of thermal design flow (85025 gpm), due to 28X tube plugging results in a 50'F increase of clad temperature which is still more than 400'F below the peak allowable temperature of 2700'F.CONCLUSION Increasing the value of the peaking factor, FQ from 1.93 to 2-1 75*for LOCA considerations, with steam generator tube plugging<28%provided reactor coolant flow stays>255,075 gpm, does not have a significant effect on the*'Proposed IOCA analysis requirement sets the limit on FQ at<2-125 as per Attachment A.
non>>LOCA analysis due to the co'nservative inputs and large margin to the safety limit.In addition, since the current non-LOCA analysis was done for an FO of 2.32 any reduction from this value is considered an additional safety margin.With regards to Overtemperature LT, Overpower W and thermal hydraulic safety limit, the results presented for 25K tube plugging are still valid, with the reactor coolant flow greater or equal to 255075 gpm.The arguments presented here are based on the fact that the reactor coolant flow rate is maintained within the limits set by the previous analysis for 25 percent level steam generator tube plugging (reference 1)~REFERENCES 1~Non-LOCA Accident Safety Evaluation for Higher Levels of Steam Generator Tube Plugging, for Turkey Point 3 and 4, June 1978 submitted with FPL letter L-78-242 dated July 20, 1978.2.Reload Safety Evaluation, Turkey Point Plant Unit 4 Cycle 4. t 5 I}}