ML20153F827
| ML20153F827 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/31/1988 |
| From: | Gottula R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20153F813 | List: |
| References | |
| ANF-88-118, NUDOCS 8809070500 | |
| Download: ML20153F827 (53) | |
Text
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4 ANF-88-118
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\\'g ADVANCED NUCLEAR FUELS CORPORATION MILLSTONE UNIT 2 LARGE BREAK LOCA/ECCS ANALYSIS 1
AUGUST 1988 l
1 8809070500 890901 PDR ADOCK 05000336 P
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4 ANF-88-118
. Issue Date:
8/23/88 l
N 3
P MILLSTONE UNIT 2 LARGE BREAK LOCA/ECCS ANALYSIS t
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Prepared by l
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l R. C. Gottula, Team Leader PWR Safety Analysis Licensing & Safety Engineering i
Fuel Engineering & Technical Services
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Contributors:
i T. Guidotti (Numerical Applications)
R. D. Hentzen (El International) j i
August 1988 i
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CUSTOMER DISCLAIMER EfPORTANT NOTICE 1%WOfMG CONTENTS ANO USS OF THis 00CuteENT PLEASE READ CAmtFULLY Advanced Nuc>eer Fuese Corporecon's warrant >es and recreeentatene con.
comeg me oustect metter of the coceent are thoes set fortn e me Agreement berwoon Aevenced Nuc$eer Pue6e Corporecon and the Cmmer pursuant to wrmcm thee docurnent e issued. hefi. esce as other.noe encrossty pro-veed in euen Agreement, neeer Aevenced Nuceu Fweis Corporaten not any person acteg on as temad manos any wertenty or recreeentaten, empresoec or vnomed, snin respect to me accuracy, comp 6etenees, or useuneos of tre iner-menon conteced e thee occument, or that the use of any etermaton, accaratus, memod or prves secoeed c mis document edi not etnnge pnvateey coned nonte; or assumes any haesince won resoect to the use of any emrmaton, ao-peresus, memod or procese oeoooed e trwe occument.
The mermeson conteced herem e for the soie use of Cwomer.
In orter to sweed umpearmere of ngnte of Advanced Nucwer Puees Carocraten e pesents or evermone wee may me ecruced e me eermaten coetsened e the oocument, the recoent, by de eccessence of tnre occument. agrees not to puesion or meme puesse use (e the patent Lee of me term)of suen eermaton uned so autnonged c wreeg Dy Advanced Nue:eer Pweis Corporaten or untd 4 Ret e.
(0) mortthe tonounng termeatson or espe'etson of tPG e4resa4 Agreemoet ano any estensen thereof, unwee emerwnee enorteety proveed e tee Agreement. No ngnte or licensee e or to any potente are ceegd ey me Nmioneg et tnie occu-ment.
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ANF-3145 472A (12 87)
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TABLE.0FCONTENfS t
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Section M
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1.0 INTRODUCTION
.a.
1 2.0
SUMMARY
OF RESULTS 1
3.0 ANALYSIS............................
3 i
L 3.1 Description of LBLOCA Transient 3
3.2 Description of Analytical Models................
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i 3.3 Plant Description and Summary of Analysis Parameters......
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i 3.4 Break Spectrum Results.....................
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3.5 Axial Shape Study Results 7
3.6 Exposure Study Results.....................
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3.'i Reducsd T,y, Operation Results.................
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4.0 CONCLUSION
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t S.0 REFERENCES...........................
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ANF 88 ll8 l
Page 11 l
l ilST OF TABLES
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3.1 Millstone Unit 2 System Analysis Parameters 9
3.2 Millstone Unit 2 Break Spectrum Analysis Results........
11 3.3 Calculated Event Times for 0.4 DECLG Break...........
12 3.4 Calculated Event Times for 0.6 DECLG Break...........
13 3.5 Calculated Event Times for 0.8 DECLG Break...........
14 3.6 Calculated Event Times for 1.0 DECLG Break...........
15 3.7 Calculated Event Times for 0.4 DECLS Break.........
16 3.8 Calculated Event Times for 0.8 DECLS Break...........
17 3.9 Calculated Event Times for 1.0 DECLS Break......,....
18 3.10 Summary of Results for 0.6 DECLG Limiting Break Size......
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o ANF 88-118 D,
Page 111 LISTOFFIGURE5 Fioure EER
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2.1 Allowable LHR as a Function of Axial Location 2
3.1 Normalized Power (EOC), 0.6 DECLG Break 20 3.2 Double intact Loop Accumulator Flow Rate.
I 0.6 DECLG Brt:ak 21 3.3 Single Intact Loop Accumulator Flow Rate, t
0.6 DECLG Brekk 22 l
3.4 Broken loop Accumulator Flow Rate, 0.6 DECLG Break.......
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3.5 Double intact Loop HPSI Flow Pate. 0.6 DECLG Break.......
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3.6 Single intact Loop SIS Flow Rate, 0.6 DECLG Break 25 t
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3.7 Broken loop SIS Flow Ratt, 0.6 DECLG Break...........
26 3.8 Upper Plenum Pressure during Blowdown, 0.6 DECLG Break.....
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t 3.9 Total Break Flow Rate during Blowdown 0.6 DECLG Break...
28 5
3.10 Pressurizer St:rge line Flow Rate during Blowdown, O.6 DECLG Brdak 29 i
3.11 Downcomer Flow Rate during Blowdown 0.6 DECLG Break......
30 3.12 Average Core Inlet Flow Rate during Blowdown, t
0.6 DECLG Break, X/L = 0.81 31 t
3,13 Hot Channel Inlet Flow Rate during Blowdo.vn, f
0.6 DECLG Break, X/L = 0.81 32 t
t 3.14 PCT Node Fluid Quality during Blowdown, 0.6 DECLG Break, X/L = 0.81 31 l
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l 3.15 PCT Node fluid Temperature during Blowduni,
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t 0.6 DECLG Break, X/L = 0.81 34 l
3,16 PCT Node Fuel Average Temperature duri.aq Blowdown.
0.6 DECLG Break I/L = 0.81
.................35 i
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ANF-88 118
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Page iv i
LIST OF FIGURES (Cont.)
E192rA E120 3.17 PCT Node Cladding Temperature during 81owdown, 0.6 DECLG Break, X/L = 0.01 36 3.18 PCT Node Heat Transfor Coefficient during Slowdown,
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0.6 DECLG Break, X/L = 0.81 37 3.19 PCT Node !! eat Flux during Blowdown, 0.6 DECLG Break, t'
X/L = 0.81 38 3.70 Containment P essure, 0.6 DECLG Break, X/L 0.81 39
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3,21 Upper Plenum Pressure after EORY, 0.6 DECLG Break, X/L = 0.81.......,.....
40 3.22 Downcomer hixture level after E0BY, 0.6 DECLG Break, X/L = 0.81...........................
41 3.23 Core Flooding Rate after E0BY, 0.6 DECLG Break, X/ L = 0. 81...........................
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3.24 Core Mixture Level after E0BY, 0.6 CECLG Break,
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X/L = 0.81.........
43 3.25 PCT Node Cladding Temperature after E0BY, 0.6 DECL6 Break, X/L = 0.81 44 I
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i ANF 88-133 Page 1 1.0 INTRCDUCTION 4
This document prsJents the results of a large break loss-of coolant accident (LOCA) analysis for the Millstone Unit 2 reactor.
The analysis was performed to support operation vith a mixed core or a core containing only ANF fuel.
Break spectrum calculations were performed to determine the limiting break size.
The break spectrum calculations included 0.4, 0.6, 0.8, and 1.0 double ended cold-leg guillotine (DECLG) break sizes and 0.4, 0.8, and 1.0 double ended cold leg-sp?it (DECLS) break sizes.
Calculations were also performed which considered axial power shapes, exposure, and full power operation with a reduced primary coolant temperature of up to 12*F.
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2.0 SUMARY OF RESULTS 4
The results of the analysis indicated the limiting break size was the 0.6 DECLG b*eak.
The Peak Cladding Temperature (PCT) for the limiting case was I
calculated to be 2163'F.
The PCT for a 12'F reduction in primary coolant temperature was calculated to be 2176*F.
The analysis supports full power cperation at 2754 MWt (2700 MWt plus 2%
unc6rtainty) with an average steam generator tube plugging of 23.5% with a maximum asymmetry of 5.9%.
The analysis supports assembly average exposures of up to 52,500 mwd /MTU.
The analysis also supports operation at full power, i
aith a primary coolant T,y, reduction of up to 12*F.
The analysis demonstrates that the 10 CFR 50.46(b) criteria are satisfied for the Millstone j
Unit a reacter with an axial and exposure independent LHR of 15.1 kW/f t as shown in Figure 2.1.
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20
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UNACCEPTABLE OPERATION 16-15.1 K w / f t 9
k N 14 -
kw 12 -
Z
+3 10-Ex3
.-a ACCEPTABLE OPERATION II) 8-4 l
M O
6-
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4-2-
E 9
T i
O.0 0.2 0.1 0.6 G8 1.0 yg e.
FRACTION OF ACTIVE FUEL Ill;IGIIT mm figure 2.1 Allowable LIIR as a function of Axiai Locatton w-w w-
-w-r
--w v
--+
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ANF-88 ll8 L
Page 3 l
3.0 ANALYSIS The purpose of the LBLOCA analysis is to demonstrate that the criteria stated in 10 CFR 50.46(b) are met. The criteria are:
1.
The eticulated peak fuel element <,1 adding temperature does not exceed the 2200'F limit.
2.
The amount of fuel element cladding which reacts chemically with water c
- steam does not exceed 1% of the total smount of zircaloy in the core.
3.
The clariding temperature tresient is terminated at a time when.the core geometry is still cmenable to cooling.
The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after qsenching.
4.
The core temperature is reduced and decay heat is removed for an extended period of tire, a5 required by the long-lived radioactivity remaining in the core.
Ctetion 3.1 of this report provides a description of the postulated large break loss c'-coolant transient.
Sectiori 3.2 describes the analytical models used in the analysis.
Section 3.3 provides a description of ths Millstone Unit 2 pl:nt and a summary of the system parameters used in the LOCA analysis.
Section 3.4 provides a summary of the results of the break spectrum calculations.
Section 3.5 summarizes the results of the axial power shape I
study.
Section 3.6 sumrurizes the results of the exposure analysis.
Section 3.7 surnmarizes '.he results of an analysis to support full poa.t operation with a rertuced primary coolani, temperature of 12*F.
3.1 Ee.Lqrjption of (Bleu Trgen_1.
A loss of-coolant accident (LOCA) is defined as the rupture of the Reactor Coolant System primary piping up to and inclurting a doub,e-ended guillotine break.
The limiting break occurs on the pump discharge side of a cold leg pipe.
Loss of-offsite power is assumed to occur co-incident with
y ANF-88-118 Page 4 the LOCA.
Primary coolant pump coastdown occurs co-incident with the loss of-offsite power.
Following the break, depressurization of the reactor coolant system, including the pressurizar, occurs.
A reactor trip signal occurs when i
l the pressurizer low pressure trip setpoint is reached.
Reactor trip and scram are conservatively neglected in the LOCA analysis.
Early in the blowdown, the reactor core experiences flow reversal and stagnation which causes the fuel rods to pass through critical heat flux (CHF).
Following CHF, the fuel rods dissipate heat through the transition and film boiling modes of heat transfer.
Rewet is precluded during blowdown by Appendix K of 10 CFR 50.
i A Safety Injection System (SIS) signal is actuated when the appropriate seisoint (high containment pressure) is reached.
Due to loss of offsite I
power, a time delay for startup of diesel generators and SIS pumps is assumed.
Once the time delay criteria is met and the system pressure falls below the l
shutoff head of the High Pressure Safety injection (HPSI) pumps and Low I
Pressure. Safety Injection (LPSI) pumps, SIS flow is injected into the cold i
legs.
Single failure criteria is met by assuming that one diesel fails.
This results in the loss of one HPSI pump and one LPSI pump.
When the system pressure falls below the Safety injection Tank (SIT) pressure, flow from the Safety Injection Tanks is injected into the cold legs.
Flow from the l
l Emergency Core Cooling System (ECCS)is assumed to bypass the core ar.4 flow to l
l the break until the end of bypass (E0BY) is predicted to occur (sustained l
downfiow in the downcomer).
Following E0BY, ECCS flow f. ills the downcomer and
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lower plenum until the liquid evel reaches the bottom of the core (beginning-l of core-recovery or BOCREC time).
During the refill period, heat is transferred from the fuel rods by radiation heat transfer.
i The reflood period begins at BOCREC time.
ECCS fluid fills the downcomer and provides the driving head to move coolant through the core.
As the l
milture level moves up the core, steam is generated.
Steam binding occurs as the steam flows through the intact and broken loop steam generators and pumps.
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The p9mps are assumed to have a locked rotor (per Appendix K of 10 CFR 50)
I which tends to reduce the reflood rate.
The fuel rods are eventually cooled j
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ANF-88-ll8 Page 5 and quenched by radiation and convective heat transfer as the quench front moves up the core.
The reflood heat transfer rate is predicted through experimentally determined heat transfer and carry-over rate fraction correlations.
3.2 Descrietion of Analytical Models The ANF EXEM/PWR evaluation model II) was used to perform the analysis.
This evaluation model consists of the following computer codes:
(1)
R00EX2(2) for computation of initial fuel stored energy, fission gas release, and gap conductance; (2)
RELAP4-EM for the system and hot channel blowdown calculations; (3) CONTEMPT /LT-22 as modi fied in accordance with NRC Branch Technical Position CSB 6-1 for computation of containment back pressure; (4) REFLEX for computation of system reflood; and (5) T000EE2 for the calculation of fuel rod heatap during the refill and reflood portions of the LOCA transient.
The quench
- time, quench velc:ity, and carryover rate fraction (CRF) correlations in REFLEX, and the heat transfer correlations in T00DEE2 are based on ANF's Fuel Cooling Test Facility (FCTF) data.
I The governing conservation equations for mass, energy, and momentum transfer are used along with appropriate correlations consistent with Appendix K of 10 CFR 50.
The reactor co'e in RELAP4 is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback, and t31th actinide and decay heating as required by Appendix K.
Appropriate conservatisms specified by Appendix K of 10 CIR 50 are ir.corporated in all of the EXEM/PWR models.
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3.3 Plant Descriotion and Summary of Analysis Parameters The Millstone Unit 2 nuclear power plant is a Combustion Engineering (CE) designed pressurized water reactor which has two hot leg pipes, two U tube f
steam generators, and four cold leg pipes with one recirculation pump in each cold leg.
The plant utilizes a large dry containment.
The reactcr coolant i
s.vstem is nodalized into control volumes representing reasonably homogeneous
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regions, interconnected by flow paths or "junctions".
The two cold legs
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l connected to the intact loop steam generator were assumed to bo symmetrical and were modeled as one intact cold leg with appropriately scaled input.
The model considers four Safety Injection Tanks, a pressurizer, and two steam' gener.itors with both primary and secondary sides of the steam generators 3
modeled.
The HP31 and LPSI pumps were mcdried as fill junctions at the SIT lines, with conservative flows given as a function of system back-pressure.
The pump performance curves were characteristic of CE pumps.
The reactor core was modeled radially with an average core and a hot assembly as carallel flow I
channels, each with three axial nodes.
A steam generator tube plugging level
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of 23.5% was assumed with a maximum asymatry of 5.9%.
The break was i
conservatively assumed to have occurred in the most highly plugged loop since this results in more steam binding during reflood and a higher peak cladding I
temperature.
Values for system parameters used in the analysis are given in Table 3.1.
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3.4 Break Soectrum Results Calculations were performed for 0.4. 0.6, 0.8 ar.d 1.0 DECLG break sizes and 0.4, 0.8, and 1.0 DECLS break sizes with an axial power shape peaked at a relative core height of 0.6 to determine the limiting break.
The break spectrum calculations were performed at a peak LHR of 15.6 kW/ft.
Beginning-of-Cycle (B0C) stored energy (where rAximum densification occurs at a hot rod average burnup of about 2000 mwd /MTU) was conservatively used in all of the break spectrum calculations.
Calculations were performed through the I
ANF 88 118 Page 7 blowdown, refill, and reflood periods of tha LOCA transient for all break d
sizes.
The results of the break spectrum study are shown in Table 3.2.
Calculated event times for the various break sizes are given in Tables 3.3 through 3.9.
The results indicated the 0.6 DECLG break size to be the l
limiting. break.
The peak cladding temperature (PCT) for the 0.6 DECLG break l
with an axial power shape peaked at a relative core height of 0.6 was l
calculated to be 1892*F. Thus, a maximum LHR of 15.6 kW/ft is supported up to l
a relative core height of 0.6.
3.5 Axial Shane Study Results Axial shape and exposure studies were performed at the limiting 0.6 DECLG j
break size. Conservative combinations of stored energy and axial power shapes
-ve ana'lyzed to bound the effects of axial power shape and exposures from BOC to EOC, In addition to the axial shape which was used in the break spectrum j
study, peaked at a relative core height of 0.6, bounding axial shapes j
representative of MOC and EOC were determined from projected axial shapes for equilibrium and xenon oscillation transient condition:;.
Two conservative i
. combinations of stored energy and axiai shape were analyzed.at a peak LHR of f
15.1 kW/f t.
They are (1) a BOC stored energy (where maximum densification l
occurs),:ombined with a MOC shape peaked at a relative core height of 0.81, and (2) a MOC stored energy combined with an EOC shape peaked at a relative i
core height of 0.9.
The results of these two calculations are shown in Table 3.10.
Case I was determined to be the limiting case with a PCT of 2163'F.
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j These calculations support a peak LHR of 15.1 kW/ft that is independent of j
axial position and exposure over the exposure range of BOC to E0C.
Plots of parameters depicting the calculations for case 1 at the limiting 0.6 DECLG f
break size are shown in Figures 3.1 through 3.25.
i 3.6 Excosure Study Results i
i The calculations described in Section 3.5 support exposures out to EOC.
l An additional calculation was performed to support assembly average exposures up to 52,500 Mdd/MTV.
This one additional calculation was sufficient to i
support assembly average exposu.es up to 52,500 mwd /MTV for the following i
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ANF-88-ll8 Page 8 reasons.
Fuel rod stored energy decreases with exposure (after peak stored energy occurs at maximum densification near BOC) out to exposures of about 40,000 mwd /MTU due to closure of the pellet cladding gap.
Beyond exposures of I
about 40,000 mwd /MTV, the stored energy increases again due to fission gas release to the gap, but is still significantly less than the stored energy at MOC.
1he rod internal pressure increases beyond exposures of about 20,000 l
mwd /MTU which tends to decrease the time of cladding rupture and also increase the pellet cladding gap which retards heat release during blowdown.
- Thus, stored energy and rod internal pressure at an exposure of 52,500 mwd /MTU bound exposures between HOC and 52,500 mwd /MTV.
The PCT calculated at an assembly l
exposure of 52,500 mwd /MTV was compared to the PCT for Case 2 described in i
Section 3.5 to verify that the PCT for exposures greater than EOC are less than PCTs calculated for exposures from BOC to EOC. The PCT at an exposure of
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52,500 mwd /MTU was calculated to be 1839'F.
Therefore, a peak LHR of 15.1 i
kW/ft is supported for assembly average exposures up to 52,500 mwd /MTV.
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3.7 Reduced T,y, Ooeration Results l
Calculations were performed to support full power operation with a reduced primary coolant temperature of up to 12'F.
A 12'F reduction in primary temperature has a small effect on the blowdown hydraulics and a small
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adverse effect on containment pressure, reflood rate, and PCT.
The lower l
initial coolant enthalpy results in a slightly lower containment pressure which results in a lower reflood rate and a higher PCT.
Calculations were i
perfermed to estimate the reduced containment pressure and its effect on reflood rate and PCT, Calculations were performed for the limiting case (0.6
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DECLG break, BOC stored energy with an MOC axial power shape) to determine the change in PCT due to the reduced primary coolant temperature.
The PCT was calculated to be 2176'F.
All 10 CFR 50.46(b) criteria are met.
A reduction in primary coolant temperature greater than 12'F is acceptable as long as there is a corresponding decrease in power level,
ANF-88-ll8 Page 9 Table 3.1 Millstone Unit 2 System Analysis Parameters Primary Heat Output, MWt 2700*
Primary Coolant Flow Rate, Ibm /hr 1.28 x 108 (340,000 gpm)
Prima *y Coolant System Volume, ft3 10,510 Operating Pressure, psia 2250 Inlet Coolant Temperature, 'F 549 Reactor Vessel Volume, ft3 4538 Pressurizer Total Volume, ft3 1500 Pressurizer Liquid Total, ft3 800 Sli Total Volume, Ft3 (one of four) 2019 SIT Liquid Volume, ft3 1150.5 SIT Pressure, psia 238.5 SIT Fluid Temperature, 'F 106.8 Tetal Number of Tubes per Steam Generator 8519 Steam Generator Tube Plugging 29.4 17.6% r.plit Number of Tubes Plugged (29.4% SGTP) 2500 (Broken Loop)
Number of Tubes Plugged (17.6% SG1P) 1500 (Double Intact Loop)
Steam Generator Secondary Side Heat Transfer Area, 29,4% SGTP, ft2 63,370 Steam Generator Secondary Side Heat Transfer Area, 17.6% SGTP, ft2 73,898 Steam Generator 3econdary Flow Rate, Ibm /hr 5.75 x 106 (29.4% SGTP)
(48 52% power split) 6.233 x 106 (17.6% SGTP)
Steam Generator Secondary Pressure (brokanloop), psia 830.1 Steam Generator Secondarv Pressure (intact loop), psia
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850.4 Steam Generator Feedwater Temperature. 'F 435 Primary Heat Output used in RELAP4-EM Model 1.02 x 2700 = 2754 MWt.
Includes pressurizer Ntal volume ar.d 23.5% average SGTP.
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Table 3.1 Millstone Point 2 System Analysis Parameters (Cont.)
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Reactor Coolant Pump Rated Head, ft 271.8 l
Raactor Coolant Pump Head, ft (DIL) 230.38*
i Reactor Coolant Pump Head, ft (SIL,BL) 233.0*
Reactor Coolant Pump Rated Torque, ft-lbf 31,560 i
Reactor Coolant Pump Rated Speed, rpm 892 l
Initial Reactor Coolant Pump Speed, rpm 874.2*
l Reactor Coolant Pump Moment of Inertia, lbm ft2 leo, coo Containment Voluine, f t3 1.938 x 106 j
rantainment Temperature, 'F
!08.1 l
$15 Fluid Temperaturo, 'F 72.8 l
11 PSI delay Time, sec 30.0 i
LPSI Delay Time, sec 50.0 l
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l Values used in RELAP4 for initialization.
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Table 3.2 Millstone tintt 2 Break Spectrum Analysis Results 0.4 DECLG 0.6 DECLG 0.8 DECtG 1.0 DECLC 0.4 DECL5 0.8 DECLS 1.0 DECLS UL-0. 6 UL-0.6 UL-0. 6 UL-0.6 UL-0.6 UL-0. 6 UL-0. 6 Peak LHR (LW/ft) 15.6 15.6 15.6 15.6 15.6 15.6 15 6 Hot Rod Burst
- Ilme (sec) 49.31 40.75 39.36 39.53 57.44 42.55 40.32
- Elevation (ft) 6.97 7.72 7.72 6.97 7.72 7.72 7.72
- Chans.el Blockage Fraction 0.25 0.3 0.26 0.25 0.3 0.3 0.28 Peak Cladding Temper & tere
- Temperature (*f) 1815.8 1892.3 1881.7 1869.6 1685.2 1761.2 1787.5
- Time (sec) 72.01 5f.15 SC.36 66.03 71.09 65.55
- 64. 42
- Elevation (ft) 7.72
- 72 7.72 7.72 1.97 7.97 7.97 Pktal Water Reaction
. tocal Maximum (%)
1.34 2.53 2.44 1.71 0.90 1.30 1.46
. Elevation of local 6.97 7.72 7.72 6.97 7.72 7.72 7.72 Man. (ft)
- Hot Pin fotal (%)
0.44 0.52 0.52 0.51 0.26 0.35 0.38
- Core Maxinum (%)
<!.0*
(1.0*
<!.0*
<l.0*
<l.0*
<!.G*
<l.0*
N
,7'
&8
- At 200 seconds.
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ANF 88 118 a
Page 12 l
1 Table 3.3 Calculated Event Times for 0.4 DECLG Break i
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[y,ggi Tima fSec) j Start 0.0 l
Break is Fully open 0.05 l
Safety injection Signal 0.90 t
Pressurizer Empties 9.19 I
SIT Injection Begins Broken Loop 20.2 Sli Injection Beginw. Single Intact Loop 21.6 t
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$1T Injection Begins. Double Intact Loop 21.6 j
End of-Bypass (E0BY) 26 !!
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Start af Reflood 39.7 l
f S!T mpties. Broken loop 57.8 i
$1T e'ppties. Double intact Loop 57.95 l
t S!T empties, Single intact Loop 58.75 j
Peak Cladding Temperature is Reached (X/L.0.6) 72.01 i
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f.NF 88 ll8 Page 13 Table 3.4 Calculated Event Timet for 0.6 DECLC Break 9
Eggat Time (Sec1 Start 0.0 Break is Fully Open 0.05 Safety Injection Signal 0.73 Pressurizer Empties 8.6 SIT Injection Begins, Broken loop 14.7 SIT Injection Begins, Single Intact Loop 16,75 SIT Injection Begins, Double Intact Loop 16.75 End of Bypass (E0BY) 21.05 Start of Reflood 34.73 S!T empties, Broken loop 52.0 Sli empties, Double intact Leop 53.0 SIT empties, Single intact Loop 53.4 Peak Cladding Temperature is Reached (X/L=0.6) 50.15
ANF 88 ll8 Page 14 Table 3.5 Calculsted Event Times for 0.8 DECLG Break
[ tant Time (See) i Start 0.0 Break is Fully Open 0.05 Safety injection Signal
. 63 Pressurizer Empties 8.6 SIT Injection Begins, Broken Locp 11.25 4
S!T Injection Begins, Single intact Loop 14.95 i
SIT Injection Begins Doubits intact Loop 14.95 End of Bypass (E0BY) 18.36 Start of Reflood 32.54 SIT empties, Broken Loop 49.45 i
Sli empties, Double intact Loop 51.35 f
SIT empties. Single Intact loop 51.6 Peak Cladding Temperature is Reached (X/L 0.6) 56.36 j
f l
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ANF 88 ll8 Page 15 Table 3.6 Calculated Event Times for 1.0 DECLG Break f1t.01 Titre (Sec)
Start 0.0 Break is Fully Open 0.05 l
Safety injection Signal 0.58 Pressurizer Empties 8.6 5!T Injection Begins, Brcken Loop 8.85 SIfInjectfonBegins,SingleIntactloop 14.5
]
SIT Injection Begins, Double Intact Loop 14.5 18.33 End of Bypass (E0BY)
Start of Reflood 32.04 SIT emptics, Broken Loop 47.8 SIT empties, Double Intact Loop 50.85 51.1 Sif empties, Single intact Loop Peak Cladding Temperature is Reached (X/L=0.6) 66.03
]
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ANF 88 ll8 Page 16 Table 3.7 Calculated Event Times for 0.4 DECLS Break
[gini Time (Sec)
Start 0.0 Break is Fully Open 0.05 Safety injection Signal 0.77 Pressurizer Empties 9.1 SIT Injection Begins, Broken loop 18.45 S!T Injection Begins, Single Intact Loop 18.6 SIT Injection Begins Double Intact Loop 18.6 End of-Bypass (E0BY) 22.29 Start of Reflood 36.r" Sif erptics, Double Intact loop
.it i
SIT empties, Single Intact loop
'A SIT empties, Broken Loop so.o Peak Cladding Temperatute is Reached (X/L 0.6) 71.09 l
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Page 17
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Table 3.8 Calculated Event Times for 0.8 D'ECLS Break i
inni T1me isul Start 0.0 Break is Fully Open 0.05 l
i Safety injection Signal 0.51 Pressurizer Empties 8.6
[
$1T Injection Begins, Broken Loop 11.25
[
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t Er.d of Bypass (E0BY) 15.0S Start of Reflood 23.77 SIT empties, Broken loop 49.05 SIT empties Double Intact Loop 49.15
$1T empties, Single intact Loop 49.35 Peak Cladding Temperature is Reached (X/L=0.6) 65.55 l
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Page 18 i
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t Table 3.9 Calculated Event Times for 1.0 DECLS Break-l
[y,tal Time (tae) i i
l Start 0.0 l
Break is Fully Open 0.05 l
Safety Injection Signal 0.5
{
Pressurizer Emptics 8.6 l
SIT Injection Begins, Broken Loop 98 Sli Injection Begins, Single Intact loop 11.75 l
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(
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l SIT empties, llouble IntNt Loop 48.85 Sli empties, Single intact Loop 49.05 l
Peak Cladding Temperature is Reached (X/L 0.6) 64.42
}
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Table 3.10 Summary of Results for b.t, DECLG Limiting Break Size 1
60,000 mwd /MTU BOC Stored Energy BOC Stored Energy MOC Stored Energy Hot Rod.
BOC Axial Shape M0C Axial Shape EOC Axial Shape Average Burn 9p
_ X/l - 0.6 X/L - 0.81 XfL - 0.9 X/L - 0.9 Peak LHR (kW/ft) 15.6 15.1 15.1 15.1
!Iot Rod Burst
- Time (sec) 40.75 35.55 37.15 46.95
- Elevation (ft) 7.72 9.22 10.22 9.97
- Channel Blockage Fraction 0.3 0.28 0.33 0.4 Peak Cladding Te.mperature
- Temperature (*F) 1892.3 2163.3 2108.0 1839.2
- Time (sec) 58.15 61.75 67.05 291.05
- Elevation (ft) 7.72 9.22 10.22 10.72 Metal-Water Reaction
- Local Maximum (%)
2.53 5.84 5.73 2.18
- Elevation of Local Max. (ft) 7.72 9.22 10.22 10.72
- Hot Pin Total (%)
0.52 0.57 0.82 0.24
- Core Maximum (%)
<l.0*
<l.0*
<l.0*
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25 2En At 200 seconds.
N$
55 At 350 seconds.
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4.0 CONCLUSION
S The results of the large break LOCA analysis for Millstone Unit 2 showed the 0.6 DECLG break size to be the limiting break with current EXEM/PWR models. The analysis supports operation (! Millstone Unit 2 at a power level of 2700 MWt and an average steam generator tube plugging level of 23.5% with a.
maximum asymmetry of 5.9%.
The analysis supports a peak LHR of 15.1 kW/ft with an axial and exposure independent power peaking limit as shown in Figure 2.1.
The analysis supports assembly average exposures of up to 52,500 mwd /MTV. The analysis supports full power operation with a reduced primary coolant temperature of up to 12'F.
A reduction in primary coolant temperature greater than 12*F is acceptable as long as there is a corresponding decrease in power level. The analysis supports Cycle 10 operation and is intended to support operation for future cycles.
Operation of Millstone Unit 2 with ANF 14x14 fuel at or below the LHR limit shown in Figure 2.1 assures that the NRC acceptance criteria (10 CFR 50.46(b)) for loss of Coolant Accident pipe breaks up to and including the double-ended severance of a reactor coolant pipe will be met with the emergency core cooling system for Millstone Unit 2.
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ANF-88-ll8 Page 46
5.0 REFERENCES
(1) Letter, Dennis H. Crutchfield (USNRC Asst. Director division of PWR Licensing-B) to Gary H. Ward (ENC Hanager, Reload Licensing), "Safety
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Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEH/PWR and Acceptance for Referencing of Re'ated Licensing Topical Reports", dated July 8, 1986.
(2) XN-NF 81-58(P)(A). Revision 1. and Sucolements 1-4, "RODEX2: Fuel Rod Thermal Mechanical Response Evaluation Model." Exxon Nuclear Company, February 1983.
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