ML20209E851
| ML20209E851 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/31/1986 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20209E842 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM TAC-48179, TAC-63133, TAC-63198, NUDOCS 8609110369 | |
| Download: ML20209E851 (40) | |
Text
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Docket No. 50-336 i
Attachment Millstone Nuclear Power Station, Unit No. 2 Item II.K.3.31, NUREG 0737 Small-Break LOCA Evaluation Large-Break LOCA Evaluation August,1986
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8609110369 860829 PDR ADOCK 05000336 P PDR
LOSS OF COOLANT ACCIDENTS RESULTING FROM PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY I. INTRODUCTION The Acceptance Criteria for LOCA analysis is described in 10CFR50.46 as follows:
- 1. The calculated fuel element peak clad temperature is below the requirement of 2200*F.
- 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, l
- 3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17 percent are not exceeded during or after quenching.
4 The core remains amenable to cooling during and after the break.
- 5. 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.
These criteria were established to provide significant margin in Emergency Core Cooling System (ECCS) performance following a LOCA.
II. MATHEMATICAL MODEL The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10CFR50.
III. SMALL BREAK LOCA ANALYSIS 4
III.1 Method of Analysis The analysis performed for Millstone Unit 2, a combustion
- Engineering NSSS, is an extension of the small break model used for Westinghouse plants. In July of 1984 the initial analysis for a generic'CE-plant, using the NOTRUMP and LOCTA computer codes for a spectrum of small breaks was performed and documented in Reference 1. Millstone 2 was the base plant used in Reference 1. The spectrum of breaks consisted I
of 3, 4, and 6-inch break analyses. It was determined that the 4-inch cold leg pump dincharge break case resulted in the highest peak clad temperature. Therefore this is the limiting i
break and the break size used in this analysis. The analysis was performed assuming 23.4 percent of all the steam generator tubes are plugged.
The NOTRUMP computer code is a one-dimensional general network code consisting of a number of advanced features, including the calculation of thermal nonequilibrium in all fluid volumes, flow regime-dependent drift flux calculations with ceanter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, and regime-dependent heat transfer correlations. NOTRUMP includes the representation of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculations. The multinode capability of the program enables an explicit and detailed spatial representa-tion of various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant transient (Reference 2 and 3 ) .
Cladding thermal analyses are performed with the LOCTA-IV (Reference 4) code which uses the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the NOTRUMP hydraulic calculations, as input.
Table 1 lists important input parameters,and initial conditions used in NOTRUMP analysis. For these analyses, the SI delivery considers pump injection flow which is presented in Table Ib as a function of RCS pressure. Minimum safeguards Emergency Core Cooling System capability and operability has also been assumed in this analysis. The core axial power distribution is shown in Figure 1.
The hydraulic analyses are performed with the NOTRUMP code using 102 percent of the licensed NSSS core power of 2700 mwt.
The core thermal transient analyses are performed with the LOCTA-IV code using 102 percent of licensed NSSS core power.
III.2 Results The results of the small break LOCA analysis are shown in Table 2 and Table 3. Figures 2a through 2h present NOTRUMP and LOCTA plots for the following key parameters of a small break ECCS analysis:
(a) Coolant Pressure in the Reactor Core (psia)
(b) Reactor Core Mixture Elevation (ft)
(c) Reactor Core Liquid Flow (lbm/sec)
(d) Total System Mass (lbm)
(e) Break Flow (lbm/sec)
(f) Hot Spot Vapor Temperature (degrees F)
(g) Fuel Rod Heat Transfer Coefficient (BTU /hr-ft2_oy)
(h) Hot Spot Clad Temperature (degrees F)
It should be noted that in some of the figures there are some oscillations or sudden steps at about 1600 seconds. This is due to injection of water from the safety injection tanks that is relatively cold compared to the vapor temperature in the core.
III.3 Conclusions An analysis of the worst case small break LOCA demonstrates acceptable ECCS performance. This analysis assumed that 23.4 percent of the steam generator tubes were plugged and a minimum RCS flow rate of 335,000 GPM. The results of this limiting case show a peak clad temperature of 2135'F and a peak local clad oxidation of 10.7 percent.
IV. LARGE BREAK LOCA ANALYSIS IV.1 Method of Analysis The current large break LOCA analysis (Reference 5) assumes 15.3 percent of the steam generator tubes are plugged and a RCS flow rate of 350,000 GPM. The current analysis has been redone assuming 23.4 percent tube plugging and a reduced RCS flow of 335,000 GPM. This analysis also includes the effects of removal of the thermal shield, adding additional steel in containment and updating of the safety injection flows to assume 75 percent HPSI, 50 percent LPSI, and 50 percent charging. The safety injection flow rates assumed for the large break LOCA are the same as those used in the small break LOCA analysis (Table 1B).
The description of the various aspects of the Westinghouse LOCA analysis methodology is given in Reference 6. This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI, WREFLOOD, COCO, and LOCTA-IV codes which are used in the LOCA analysis are described in detail in References 7 through 10; code modifications are specified in References 11 through 17.
These codes are used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdown, and the WREFLOOD computer code is used to calculate the containment pressure transient throughout the LOCA analysis. The LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the entire analysis.
SATAN-VI is used to calculate the RCS pressure, enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the
blowdown phase of the LOCA. SATAN-VI also calculates the safety injection tank water flow rates and internal pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown. At the end of the blowdown phase, these data are transferred to the WREFLOOD code. The mass and energy release rates during blowdown are utilized in the COCO code for use in the determination of the containment pressure response during this first phase of the LOCA. Additional SATAN-VI output data including the core flow rates and enthalpy, the core pressure, and the core power decay transient, are transferred to the LOCTA-IV code.
With initial information from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the core water level during the refill and reflood phases of the LOCA. WREFLOOD also calculates the mass and energy flow addition to the containment through the break. Since the mass flow rate to the containment depends upon the core flooding rate and the local core pressure, which is a function of the containment backpressure, the WREFLOOD and COCO codes are interactively linked. WREFLOOD is also linked to the LOCTA-IV in its calculation of the fuel temperature. LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel clad temperature and metal-water reaction of the hottest rod in the core.
The analysis presented here was performed with the 1981 version of the evaluation model which includes modifications in response to NUREG-0630 (Reference 17). Reactor coolant pumps are assumed to continue to run during blowdown unless otherwise noted.
IV.2 Results 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 5. 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 4 presents the occurrence time for various events throughout the accident.
Table 5 presents selected input values and results from the hot fuel rod thermal transient calculation. For these i
results, the hot spot is defined as the location of maximum peak clad temperatures. That location is specified in Table 5 for the worst break case analyzed. The location is
indicated in feet which presents elevation above the bottom of the active fuel stack.
Table 6 presents a summary of the various containment systems parameters and structural parameters which were used as input to the COCO computer code (Reference 9) used in this analysis.
Figures 3a through 3n present the parameters of principal interest from the large break ECCS analysis. The following items are noted:
Figure 3a Hot spot clad temperature Figure 3b Coolant pressure in the reactor core Figure 3c Water level in the core and downcomer during reflood Figure 3d Containment pressure transient Figure 3e Core flow during blowdown Figure 3f Fuel rod heat transfer coefficients Figure 3g Hot spot fluid temperature Figure 3h Mass released to containment during blowdown Figure 3i Energy released to containment during blowdown Figure 3j Fluid quality in the hot assembly Figure 3k Mass velocity Figure 31 Safety Injection Tank water flow rate into RCS during blowdown (per tank)
Figure 3m Pumped safety injection water flow rate during reflood Figure 3n Core reflooding rate IV.3 Conclusions For breaks 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 Section 1.
The results of this limiting case show a peak clad temperature of 2142'F and a peak local clad oxidation of 6.17 percent.
V. REFERENCES
- 1. Bajorek, S. M., " Addendum to the Westinghouse Small Break ECCS Evaluation Model (WCAP-10054) using the NOTRUMP code for the Combustion Engineering NSSS,"
WCAP-10054-addendum 1, July 1984.
- 2. Lee, H., Rupprecht, S. D., Tauche, W. D., Schwraz, W. R.,
" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, August 1985.
- 3. Meyer, P. E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A, August 1985.
- 4. Bordelon, F. M., et al., "LOCTA-IV Program:
Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary), and WCAP-8305 (Nonproprietary),
June 1974.
- 5. Letter from W. G. Counsil to R. A. Clark, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revisions to Technical Specifications, Large and Small Break Loss-of-Coolant Accident Evaluation," October 22, 1982.
- 6. Bordelon, F. M., Massie, H. W., and Zordan, T. A.,
" Westinghouse ECCS Evaluation Model - Summary,"
WCAP-8339, July 1974.
- 7. Bordelon, F. M., et al., " SATAN-VI Program:
Comprehensive Space-Time-Dependent Analysis of Loss of Coolant," WCAP-8302 (Proprietary) and WCAP-8306 (Nonproprietary), June 1974.
- 8. Kelly, R. D., et al., " Calculation Model for Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Nonproprietary), June 1974.
- 9. Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (COCO)," WCAP-8327 (Proprietary) and WCAP-8326 (Nonproprietary), June 1974.
- 10. Bordelon, F. M., et al., "LOCTA-IV Program:
Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Nonproprietary), June 1974.
- 11. Ferguson, K. L., and Kemper, R. M., "ECCS Evaluation Model for Westinghouse Fuel Reloads of Combustion Engineering NSSS," WCAP-9528 (Proprietary) and WCAP-9529 (Nonproprietary), June 1979.
- 12. Ferguson, K. L., and Kemper, R. M., Addendum to ECCS Evaluation Model for Westinghouse Fuel Reloads of Combustion Engineering NSSS," October 1979.
_7-
- 13. Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model - Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Nonproprietary), April 1975.
- 14. " Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Nonproprietary), November 1975.
- 15. Letter NS-CE-924, dated January 23, 1976, C. Eicheldinger (Westinghouse) to D. B. Vassallo (NRC).
- 16. Eicheldinger, C., " Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Nonproprietary Version),
February 1978.
- 17. " Westinghouse ECCS Evaluation Model - 1981 Version,"
WCAP-9220-P-A Revision I (Proprietary), WCAP-9221-A Revision I (Nonproprietary), February 1982.
TABLE 1
+ INPUT PARAMETERS USED IN THE SMALL BREAK ANALYSES Parameter Small Break Peak Linear Power (kw/ft) 15.21 (includes 102 percent factor)
Total Peaking Factor, F g 2.464 Power Shape See Figure 1 Fuel Assembly Array 14 x 14 T 1080 Nominal Water Cold Volume Leg(ftSj/ accumulator)
Nominal Cold Leg SIT 2019 3
Volume (ft / accumulator)
Minimum Cold Leg SIT 215 Gas Pressure (psia)
Pumped Safety Injection Flow See Table Ib Stea.a Generator Tube Plugging Level 23.4%
Steam Generator Initial Pressure (psia) 810.0 RCS Flow (gpm) 335000 Cold leg Temperature (steady state) 551.5 (degrees F)
F TABLE 1b SAFETY INJECTION FLOW RCS Pressure (psia) SI Flow (gpm) 14.7 2073.00 40.0 1961.25 80.0 1750.25 120.0 1515.25 160.0 1157 00 200.0 647 25 209.0 445.00 '
300.0 422.50 400.0 400.00 500.0 378.25 600.0 348.25 700.0 326.50 800.0 300.25 900.0 266.50 1000.0 226.00 1100.0 170.50 1220.0 81.25 1225.0 22.00 1800.0 22.00 The assumptions made for the SI flow are 75% HPSI, 50% LPSI, and 50% charging flow O
I i
, - . - - - . - . , . , , . - _ _ .__-.,.,,,..---..,,,,._-_.n - _ . _ - , - . - - _ - - . - - _ - - - - - . _ , . - - - - . . +
_ .. _ ..___ . _. - _ . . - . _=- . . _ . _ . _ _ _ _ _
L TABLE 2 SMALL BREAK LOCA TIME SEQUENCE OF EVENTS Time (sec) l Start 0.0 Reactor Trip 7.04
, Top of Core Uncovered 877.0~
SIT Injection (start) 1600.0
- SIT Injection (stop) 1658.0 s
Peak Clad Temperature Occurs 1659.0 1
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0 0 TABLE 3 SMALL BREAK LOCA RESULTS FUEL CLADDING DATA I
Peak Clad Temperature (*F) 2135 Peak Clad Location (ft) 11.14 Local Zr/H O Reaction (max, %) 10.70 2
Local Zr/H2 O Reaction Location (ft) 11.14 Total Zr/H 2O Reaction (%) <.3 Hot Rod Burst Time (sec) 1481.0 i
Hot Rod Burst Location (ft) 11.14 i
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3 TABLE 4 i
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LARGE BREAK TIME SEQUENCE OF EVENTS l
- CD =0.6 DECW i
_ (sec)
START 0.0 l S..I. Signal
- 0.65 l SIT Tank Injection 15.6 4
End of Blowdown 21.6 i
Bottom of Core Recovery 34.7 i
j SIT Tank Empty 64.4 4
l End of Bypass 21.6 Time of PCT 169.4 4
4 l
7
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i Table 5 LARGE BREAK Results I D=0.6 D m n Peak Clad Temp. , F 2142 Peak Clad Location, ft. 7.5 Local Zr/H 2O Rxn(max), 5 6.17 Local Zr/H2 O Location, ft. 7.5 Tctal Zr/H 2O Rxn, 5 <03 Hot Rod Burst Time, sec. 25.6 Hot Rod Burst Location, ft. 5.46 Total Peaking Factor, Fg 2.464 NSSS Power, Nt,102% of 2700 Peak Core Linear Power, kw/ft 15.6 Accumulator Tank Actuation Pressure, psia 215 Accumulator Tank Water Volume, 3ft per tank 1080 Ptaped Safety Injection See Table Ib
Table 6 j Millstone Unit 2 Containment Physical Parar.eters 6
Net Free Volume 1.938 x 10 ft3 Containment Initial Conditions:
Humidity 9 Containment Temperature 6 F Enclosure Building Temperature 60 F Ground Temperature 40 F Initial Pressure 14.7 psia Initial Time for:
Spray Flcw 26 seconds Fans (3) 0.0 seconds Additjonal Fan 14.0 seconds Containment Spray Water:
Temperature 50 F
- Flcw Rate (Total, 2 pumps) 3300 gpm Fan Cooling Capacity (per fan)
Vnmr Tenernture (OF) Canacity (Etu/sec) 60 0.0 145 3360.0 165 5280.0 300 28800.0 350 32400.0 Containment Heat Absorbing Surfaces
- 1. Surface Area; and Thicknesses
- a. Shell and Dom - 71,870 ft2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.25 in.
(3) Concrete - 3 0 ft. (one side exposed to enclosure building atmosphere)
- b. Unlined Concrete - 62,800 ft 2 (1) Concrete - 2.0 ft. (one side exposed to containment atmosphere, one side insulated)
- c. Galvanized Steel - 120,000 ft (1) Zinc - 0.0036. in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.20 in. (one side insulated)
Tcble 6 (continu::d)
Millstone Unit 2 Contni runt Physical Paraneters
- d. Painted Thin Steel - 56,350 ft2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.2 in. (one side insulated)
- e. Painted Steel - 32,600 ft 2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.26 in. (one side insulated)
- f. Painted Steel - 22,425 ft 2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.86 in. (one side insulated)
- g. Painted Thick Steel - 4,230 ft 2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 2 94 in. (one side insulated)
- h. Containment Penetration t.rea - 3,000 ft2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0 75 in.
(3) Concrete - 3.75 ft. (one side exposed to enclosure building atmosphere)
- 1. Stainless Steel Line Concrete - 8,340 ft2 (1) Stainless steel - 0.25 in. (one side exposed to containment atmc:: nere)
(2) Concrete - 2.0 ft. (one side insulated)
- j. Base Slab - 11,130 ft 2 (1) Concrete - S.0 ft. (one side exposed to containment sump, one side exposed to ground)
- k. Neutron Shield - 1400 ft (1) Stainless steel - 0.024 ft. (both sides exposed to containment atmosphere)
- 1. CEIN Cable Support Structure - 1380 ft (1) Paint - 0.006 in.
(2) Stainless Steel - 0.1094 ft. (both sides exposed to containment atmosphere)
- m. Painted Steel - 1856 ft2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.25 in. (one side insulated) 2
- n. Painted Steel - 350 ft (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0.312 in. (one side insulated)
Table 6 (continusd)
Millstone Unit 2 Contai nrrent Physical Parameters
- o. Painted Steel - 2848 ft2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel - 0 375 in. (one side insulated)
- p. Painted Steel - 43 ft2 (1) Paint - 0.003 in. (one side exposed to containment atmosphere)
(2) Carbon steel .437 in. (one side insulated)
- 2. Thermal Properties Conductivi HeatCagagity Material IBtu/hr-ft gyF) (Btu /ft - F1
- a. Concrete 2.0 36
- b. . Carbon Steel 35.0 55
- c. Stainless Steel 10.0 62
- d. Paint 1.5 32
- e. Zinc 70.0 45 l
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g i e s r u
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q i F
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2
. ( b O.O 0 0 0 0 0
. 0 0 0 0 0 0 0 0 0
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0
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o~ a wxawew" +zwIz_<5zOJ L
i
' iii1 i:
MP2 LBLOCA SGTP= 23.4 CD=0.6 PUMPS RUNNING MILLSTONE-2 SATAN DECLG Z-FLOWRATE CORE BOTTOM () TOP , (*)
seee k 6969 dece e
! 2 ees y k-
, ( .
% r"-
2998 '
4896
\f
. -6888 k
8 2 4 6 8 18 12 14 16 19 28 22 TIME 15ECI e
Figure 3e Core Flow During Blowdown
a MILLSTONE-2 CD=0.6 SGTP=23.4%
PUMPS RUNNING DECLG HEAT TRANS. COEFFICIENT BURST, 5.46 FT( ) PEAK, 7.50 FT( * )
te5 4 I y I A .
I E
I/, : /
$ se?
- \ \ l 5 u G l C i i l y y - - -
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= '
E 7 i' .
ice e as se 7s see nas ise 17s 2ee 22s ase 27s see s2s TIME isECl Figure 3f Fuel Rod Heat Transfer Coefficients
r B
MILLSTONE-2 CD=0.6 SGTP=23.4%
PUMPS RUNNING DECLG FLUID TEMPERATURE BURST, 5.46. FT( ) PEAK, 7.50 FT(- )
2ses
- 1758 5
g 1589
,25.
1998
,5. . /
E f W }
See y a
5 259 2 2
- - ^ ^
- 2 --
8 25 58 75 IBO 125 158 175 208 225 256 275 585 525 TIME ISECB Figure 3g Hot Spot Fluid Temperature
e MP2 LBLOCA SGTP:23.4 CD-0.6 PUMPS RUNNING MILLSTONE-2 SATAN DECLG BREAK FLOW
.0805[*5 l 7853E*5 m
1 a .5888t 5 g .588K+5 x s \
400EE *5
.90EEE 5
.200E .5 '
4'
.lGCSE 5
- s. N
.I800C+5 O. 2. 4 6. 9. IS. 12. 14 16. 30. 20. 22.
TIT estC e i
Figure 3h Mass Released to Containment During Blowdown (lbm/sec) i I
t
~
MP2 LBLOCA SGTP:23.4 CD:0.6 PUMPS RUNNING MILLSTONE-2 SATAN DECLG BREAK ENERGY l .samet.e i
. . M *0 g .9 M 9 N 5
.J WOt=0 A l
. t .. _ \
\
\
N
- e. N
.3000t*O S. 2. 4 8. S. 18. 12. 14 16. ft. 29, 22.
tgest estCs Figure 3i Energy Released to Containment During Blowdown (BTU /sec)
c.
a MILLSTONE-2 CD=0.6 SGTP=23.4%
PUMPS RUNNING DECLG
, QUALITY OF FLUID BURST, 5.46 FT( ) PEAK, 7.50. FT(-)
16
.i ie 5
y12
?
- ^ ^ ^
1 1
O 5 / Y Y 6 l ,
b /
t a '
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h2 I e
les 3,i i,2 3,5 TINE ISECl 1
Figure 3j Fluid Quality in the Hot Assembly
A 4'
4 MILLSTONE-2 CD=0.6 SGTP=23.4%
PUMPS RUNNING DECLG I
MASS VELOCITY BURST, 5.46 FT( ) PEAK, 7.50 FT( * )
58 '
h V \
U \'
h! e .
r 0 %
s 6
$ Se 7
\
-186 f E //
- /
y 156 d O 206
//
258 18' 15 ge2 gg5 TIME (SECl
! Figure 3k Mass Velocity
i 0
! i 4
- MP2 LBLOCA SGTP= 23.4 CD=0.6 PUMPS RUNNING i
) MILLSTONE-2 SATAN DECLG 1
l ACCUM. FLOW j asse i 1 -
1 O l 'e j 3
- 2ees 3
v/
l, 3
u g1580 v i
. IM see '
e e 2 4 6 8 le 12 14 16 18 28 22 111t ISECl 1 .
Figure 31 Safety Injection Tank Water Flow Rate Into RCS During Blowdown (per tank)'
I
- 00'005 00' 00ir I e a
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o 1
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9
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(
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MP2 LBLOCA SGTP=23.4% CD=0.6 PUMPS RUNNING MILLSTONE-2 WREFLOOD/ COCO DECLG FLOOD RATE (IN/SEC) 2 t
GIG l W 1
\N 5 N w y
er y
i l m a
g ~
~
l 3
1 1
l 1
8
! 28 48 68 88 188 129 188 168 198 288 228 248 268 Ptf8 588 528 TINC ISEC1 Figure 3n Core Reflooding Rate
__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ ___