L-76-419, Letter Regarding ECCS Reevaluation
| ML18227C908 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 12/09/1976 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| L-76-419 | |
| Download: ML18227C908 (141) | |
Text
hlRC FORM 195 I2~ 76 I U.S. NUCLEAR REGULATORY COIAMISSION NRC DISTRIBUTION soR PART EO DOCKLI P MAI'ERIAL DOCK 8ER
-2 251 TO:
Hr. Victor Stello FROM:
Florida Power
& Light Company Hiami, Florida Hr. Robert E. Uhrig DATE OF DOCUMENT 12/9/76 DATE RECEIVED 12/9/76
%LETTER JHl ORIGINAL Qcopv DESCRIPTION DNOTORIZED gUNC LASSIF IE D PROP INPUT FORM ENCLOSURE NUMSER OF COPIES RECEIVED Three signed 35 copies encl recvd.
Ltr. w/attached erra ta shee t.... re our 6/17/76 and 12/3/76 orders...trans the following:
SOMOT REMOVE ACKNOWLEDGED PLANT NAME:
Turkey POIInt Units 3 & 4 (3-P)
Concerns re-evaluation of ECCS cooling performance calculated in accordance with an approved Nestinghouse Evaluation Model, with appropriate corrections for upper head water temperature.
(6O-P)
SAFETY ASSXGNED AD:
RO EC MANA L C ASST Lear Elliott arrz.s FOR ACTION/INFORMATION PROJECT MANAGER'IC ASST 12 10 76 INTERNALDISTRIBUTION REG FILE I &E OELD GOSSICK & STAFF MIPC CASE HANAUER HARLESS PROJECT MANAGEMENT BOYD PE COLLINS HOUSTON PETERSON MELTZ HELTEMES SKOVHOLT LPDR ~ Miami, Fla.
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FLORIOA POWER & LIGHTCOMPANY December 9,
1976 L-76-419 Office of Nuclear Reactor Regulation Attention:
Mr. Victor Stello, Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555 pegutatory Docket I Iie
Dear Mr. Stello:
Re:
Turkey Point Units 3
6 4
Docket Nos.
50-250 and 50-251 ECCS Re-evaluation 0
+
<-'~'~o~
+e In accordance with the June 17, 1976 and December 3,
1976 Orders for Modification of License issued by the Commission for Turkey Point Unit 4, Florida Power 6 Light Company hereby submits a
re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model, with appropriate corrections for upper head water. temperature..
The re-evaluation also includes the effects of (1) plugged steam generator tubes and (2), modifications to increase the water volume of the Safety Injection accumulators.
The re-evaluation was performed using a rated power level of 2192 Mwt instead of the actual Technical Specification rated power level of 2200 Mwt.'estinghouse Electric Corporation is now revising their calcula-tions using 2200 Mwt and, when this is complete, change pages will be forwarded to you to bring the ECCS re-evaluation up to date.
The ECCS re-evaluation assumes
-a minimum accumulator water volume of 875 cubic feet,
- however, the actual m'inimum'water volume for the remainder of core cycle 3 for Unit 4 will be 825 cubic feet.
Modifications to increase the Unit 4 accumulator volume are planned for the Spring 1977 refueling outage.
Modifications to increase the Unit 3 accumulator volume will be completed during the Fall 1976 refueling outage which is now in progress.
Core Cycle 3 parameters were used in the ECCS re-evaluat'on.
Westinghouse Electric Corporation then compared Cycle 4
operation with Cycle 3 operation and concluded that Cycle 3 is PEOPLE... SERVING PEOPLE
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Office of Nuclear Reactor Regulation Attention:
Mr. Victor Stello, Director Page Two more limiting than Cycle 4 with respect to ECCS performance.
The comparison included consideration of the three Region 3
assemblies which will be reloaded into Unit 4 for Cycle 4
operation.
Proposed Technical Specification amendments incorporating the results of the ECCS re-evaluation will be submitted under separate cover letters.
Very tru ursg
- Robert, E. Uhrig Vice President REU/MAS/cpc Attachment cc:
Mr. Norman C. Moseley Robert Lowenstein, Esquire
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ERRATA SHEET Page 1, paragraph 1, line 7:
replace the words "in Section 15.3.1" with the word "herein".
Page 1, paragraph 2, line 1:
add the word "would" after the words "Reactor Coolant System".
Page 1, paragraph 2, line 4:
the word "injection" should be capitalized.
Page 2, paragraph 3, line 2:
The word "serverance" should be spelled "severance".
Page 3, paragraph 1, last line:
replace the Table number "15.4-3" with the number "3".
Page 3, paragraph 4, lines 6 and 7:
delete the words "the value of the peak linear power density used in this analysis and".
This same phrase is used twice.
Page 4, paragraph 2, first subparagraph:
the subparagraph heading should be "Figures la through 3d".
Page 5, subparagraph 2, last line:
the word "intack" should be spelled "intact".
Page 5, main paragraph 2, line 3:
the word "reacotr" should be spelled "reactor".
Page 5, main paragraph 2, line 4:
the word "id" should be spelled "is".
Page 6, paragraph 1, line 3:
delete the period after the word "saturation".
Page 6, paragraph 1, line 6:
the word "whos" should be spelled "whose".
Page 6, paragraph 3, line 1:
the word "toal" should be spelled "total", and the number "013" should be ".013".
Page 10, Table 2, second column:
the expression "CD+0.6" should be "CD=0.6".
Page 10, Table 2, third and fourth columns:
the expressions "CD=0.4" should be completely, enclosed by parentheses.
Page 10, Table 2, near the bottom of the page:
the word "cucle" should be spelled "cycle".
Page ll, Table 3, last line:
the word "Fastests" should be spelled "Fastest".
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<a
MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT)
The analysis specified by 10CFR50.46 Acceptance Criteria for Emer-Cl]
gency Core Cooling Systems for Light Water Power Reactors",
is presented in this section.
The results of the loss of coolant accident analvses are shown in Table 2
and show compliance with the Acceptance Criteria.
The analytical techniques used are in compliance with Appendix K of
- 10CFR50, and are described in Reference
[2].
The results for the small
- break loss of coolant accident are presented in Section 15.3.1 and are in conformance with 10CFR50.46 and Appendix K of 1GCRR50.
Should a major break occur, depressurization of the Reactor Coolant System result in a pressure decrease in the pressurizer.
Reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached.
A Safety injection System signal is actuated when the appropriate set-point is reached.
These countermeasures will limit the consequences of the accident in two ways:
Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
\\
2 Injection of borated water provides heat transfer from the core and prevents excessive clad temperatures.
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At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liauid which transfers heat from the core by forced convection with some fully developed nucleate boiling.
After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10CPR50.
Thereafter the core heat transfer is based on local conditions with boiling and forced con-vection to steam as the major heat transfer transition mechanisms.
During the refill period, rod-to-rod radiation is the only heat transfer mech-anism.
When the Reactor Coolant System pressure falls below 600 psia the ac-cumulators begin to inject borated water.
The conservative assumption is made that accumulator water injected bypasses the core and goes out through the break until the termination of bypass, 'his conservatism is again consistent with Appendix K of 10CFR50.
Thermal Anal sis Westinghouse Performance Criteria for Emergency Core Cooling System.
The reactor is designed to withstand thermal effects caused by a loss of coolant accident including the double ended serverance of the largest Reactor Coolant System pipe.
The reactor core and internals together with the Emergency Core Cooling System (ECCS) are designed zo that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident.
The
- ECCS, even when operating during the injectio n mode with the most severe single active failure, is designed to meet the Acceptance Criteria December 1976
41
Method of Thermal Analysis The description of the various aspects of the loss of coolant accident analysis is given in Reference
[2].
This document describes the major phenomena
- modeled, the interfaces among the computer codes and features of the codes which maintain compliance with the Acceptance Criteria.
The individual codes are described in detail in References
[3] through
[6].
The analyses presented here were performed using the October 1975 version of the Nestinghouse Evaluation Model.
This version includes the modifications to the models, referenced
- above, as specified by the Nuclear Regulatory Commission (NRC) in Reference
[7] and complies with Appendix K, of 10CPR50.
The October, 1975 Llestinghouse Evaluation Model is documented in References
[8, ll and 12].
Containment data used to calculate ECCS backpressure is presented in Table 15.4-3.
Table 1 presents the time sequence of events for both ECCS analyses using the NRC and Westinghouse Parameters.
Results Table 2 presents the peak clad temperatures and hot spot metal reaction for a range of break sizes.
This range of break sizes was de-termined to include the limiting case for peak clad temperature from sensitivity studies reported in References 9 and 10.
The Satan VI analysis of the loss of coolant accident is performed at 102 percent of Licensed Application Core Power Level (power level shown in Table 2).
The peak linear power, and peaking factor at the license application power level'sed in the'analyses, are also given in Table 2.
Since there is margin between the value of the peak linear power density used in this analysis and the value of the peak linear power density used in this analysis and the value expected in operation, a
lower peak clad temperature would be obtained by using the peak linear power density expected during operation.
December 1976
4 A
~ 4 I
4
Three cases are analyzed with 5% uniform steam generator tube plugging.
An additional case is presented, for the limiting break, with 10%
uniform steam generator tube plugging.
For the results discussed below, the hot spot is defined to be the location of maxiumum peak clad temperature.
This location is given in Table 2 for each break size analyzed.
Figures 1 through 16 present the transients for the principal parameters for the break sizes analyzed.
The following items are noted:
Figures la The following quantities are presented at the clad burst location and at the hot spot (location of maximum clad temperature) both on the hottest fuel rod (hot rod):
(1) fluid quality (2) mass velocity (3) heat transfer coefficient.
The heat transfer coefficient shown is calculated by the LOCTA IV code.
Figures 4a The system pressure shown is the calculated pressure in the core.
The flow rate out the break is plotted as the sum of both ends for the guillotine break cases.
This core pressure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet.
Figures 7a through 9d These figures show the hot spot clad temperature transient and the clad temperature transient at the burst location.
The fluid temperature shown is also for the hot spot and burst location.
The core flow (top and bottom) is also shown.
Figures 10a These figures show the core reflood transient.
through 10d December 1976
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Figures lla through 12d These figures show the Emergency Core Cooling System flow for all cases analyzed.
.As described earlier, the accumulator delivery during'lowdown is discarded until the end of bypass is calculated.
Accumulator flow, however, is established in refill reflood calculations.
The accumulator flow assumed is the sum of that injected in the intack cold legs.
Figures 13 a,b,c,d The containment pressure transient is also provided.
Figures 14 These figures show the core power transient.
a,b,c,d Figures 15 a,b,c,d These figures show the break energy released to the containment during blowdown for the limiting case break.
Figure 16 This fC.gure provides the containment wall condensing heat transfer coefficient for the limiting case break.
In addition to the above, Tables 4 and 5 present the reflood mass and energy release to the containment and the broken loop accumulator mass and energy flowrate to the containment, respectively.
The analysis presented in this section was performed with a reactor vessel upper head temperature equal to the RCS hot leg temperature.
The effect of using the hot leg temperature in the reacotr vessel upper head id described in Reference 13.
A break spectrum sensitivity study using the hot leg temperature is presented in Reference
[14].
The purpose of the Reference 14 sensitivity study is to show that changing the upper head water temperature does not change the limiting break type and location, which is a Double Ended Cold Leg Guillotine, for a three loop plant.
The three loop plant configuration used for this sensitivity study is sufficiently similar to the Turkey Point Units 3 and 4 plants to assure that the limiting break is identified.
December 1976
'4 0
ai
Particular details of p ant design do not change the ba c effect re-sulting from the higher upper head temperature, that being the flashing of water at the hot leg saturation pressure rather than the cold leg saturation.
pressure.
In addition previous sensitivity studies (References 9, 10, and 14) performed for 3 loop plants have been consistent, in verifying this limiting break type and location for Mestinghouse plants whos designs reflect the differences noted between Reference 14 and the Turkey Point 3
6 4 plants.
The clad temperature analysis is based on a total peaking factor of 2.25.
The hot spot metal water reaction reached is 10.732%, which is well below the embrittlement limit of 17 percent, as required by 10CFR50.46.
In addition, the toal core metal-water reaction is less than 013 percent for all breaks as compared with the 1 percent criterion of 10CFR50.46.
The results of several sensitivity studies are reported in Reference 9.
These results are for conditions which are not limiting in nature and hence are reported on a generic basis.
For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Accep-tance Criteria as presented in 10CFR50.46.
That is; l.
The calculated peak fuel element clad temperature provides margin to the requirement of 2200'F based on Fq value of 2.25.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
3 ~
The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The cladding oxidation limits of 17 percent are not exceeded during or after quenching.
4.
The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radio-activity remaining in the core.
December 1976
'a
REFERENCES "Acceptance Criteria for Emergency Core Cooling Systems for Light Hater Cooled Nuclear Power Reactors,"
10CFR50.46 and Appendix K of 10CFR50.
Federal Register, Volume 39 Number 3, January 4, 1974.
Bordelon, F.M., Massie, H.H. and Z ordon T.A., "Westinghouse FCCS Evaluation Model Summary," WCAP-8339, July, 1974 Bordelon, F.M., et al.,
"SATAN-VI Program:
Comprehensive Space-Time Dependent Analysis of Loss of Coolant," HCAP-8302, June, 1974 (Proprietary) and WCAP-8306, June, 1974 (Non-Proprietary).
- Bordelon, F. M., et al.,
"LOCTA-IV Program:
Loss of Coolant Tran-sient Analysis," WCAP-8301, June 1974 (Proprietary) and HCAP-8305,
- June, 1974 (Non-Proprietary).
Kelly, R. D., et al., "Calculational Model for Core Reflooding after a Loss of Coolant Accident (HREFLOOD Code)," WCAP-8170, June 1974 (Proprietary) and HCAP -8171, June 1974 (Non-Proprietary)
- Bordelon, F.
M. and Murphy, E. T., "Containment Pressure Analysis Code (COCO)," WCAP-8327, June, 1974 (Proprietary) and HCAP-8326,
- June, 1974 (Non-Proprietary)
"Supplement to the Status Report by the Directorate of Licensing in the matter of Westinghouse Flectric Company ECCS Evaluation Model Conformance to 10CFR50 Appendix K." Federal Register,
- November, 1974.
Bordelon, F.M., et al., "Westinghouse ECCS Evaluation Model-Supplementary Information," HCAP-8471-P-A,April, 1975, (Proprietary) and HCAP 8472-A, April, 1975 (Non-Proprietary).
Salvatori, R., "Westinghouse ECCS - Plant Sensitivitv Studies,"
HCAP-8340, July, 1974 (Proprietary) and HCAP-8356, July, 1974 (Non-Proprietary).
Buterbaugh, T. L., Julian, H.V. and Tome, A.E., "Westinghouse ECCS-Three Loop Plant (17x17) Sensitivity Studies" WCAP 8572-P
- July, 1975 (Proprietary) and WCAP 8573, July, 1975 (Non-Proprietary).
December 1976
"Westinghouse ECCS Evaluation 1'fodel October, 1975 Version, WCAP 8622, November 1975 (Proprietary) and WCAP-8623, November
- 1975, (Non-Proprietary)
Letter from C. Eicheldinger of Westinghouse Electric Corporation to D. B. Vassallo of the Nuclear Regulartory Commission; letter number NS-CE-924, dated January 23, 1976.
Letter from C. Eicheldinger of Westinghouse Electric Corporation to V. Stello of the Nuclear Regulatory Commission, Letter Number NS-CE-1163 dated August 13,
- 1976, Julian, H. V., Tabone, C. J.,
and Thompson, C. N., "Westinghouse ECCS-Three Loop Plant (17x17) Sensitivity Studies, WCAP 8853, September, 1976 (Non-Proprietary)
December 1976
Ai 4,
TABLE 1 LARGE BREA'INE SEOUENCE OF EVENTS
- DECL (CD-1.0)
(Sec)
+DECL (CD=0.6)
(Sec)
~DECL (CD=0.4)
(Sec) o'ctcDECLG (CD=0.4)
(Sec)
START Reactor Trip Signal S. I'ignal 0.0 0.573 0.44 0.0 0.583 0.55 0.0 0,595 0.67 0.0 0.595 0.67 Acc. Injection End of Bypass End of Blowdown
- 10. 3 20.96 21.18 Acc. Empty Pump Injection 55.09 25.44 Bottom of Core Recovery 40.37 12 '
- 23. 81 23.96 42.78 57.71 25.55 16.6
- 27. 89 28
~ 04 46.94
- 61. 52 25.67 16.3 27.66 27.84 46.78
- 61. 19
- 25. 67
>'< Contains 5% Stm.
Gen.
Tube Plugging (Figures lc thru 14c)
Contains 10% Stm.
Gen.
Tube Plugging (Figures ld through 14d, 15 and 16)
- December, 1976
TABLE 2 L+GE BREAK Results Results
>>DECL (CD-1.0)
'<DECL
<<DF.CL (CD+0.6)
(CD=0.4
>>>>DECLG (CD=0.4 Peak Clad Temp. 'F Peak Clad Location Ft.
Local Zr/H20 Rxn(max)%
Local Zr/H20 Location Ft.
Total Zr/H20 Rxn Hot Rod Burst Time Sec.
Hog Rod.Burst Location Ft.
1767 6.25 2.009 6,0 L0.3 45.24 6
~ 0 1927 6.5 3.627 6.0 L0.3 31.0 6.0 2162 6.0 10.732 6.0 LO ~ 3 24.1 6.0 2198 6.0 12.310 6 '
L0.3 22,9 6.0 Calculation Core Power'wt 102% of Peak Linear Power kw/ft 102% of Peaking Factor Accumulator Mater Volume (ft3) 2192 2.25
~875
( ar accumulator)
Fuel region + cucle analyzed UNITS 3 and 4
Cycle Region
- Contains 5% Stm.
Gen.
Tube Pilugging
>>>> Contains 10% Stm Gen.
Tube Plugging December 1976 10
Jl 0
0
TABLE 3 LARGE BREAK CONTAINNFNT DATA (DRY CONTAINMENT)
NET FREE VOLlNi E 1.55xl0 Ft3 INITIALCONDITIONS Pressure Temperature RMST Temperature Service Hater Temperature Outside Temperature 14.7 psia 900&
39'F 63'F 39'F SPRAY SYSTEM Number of Pumps Operating Runout Flow Rate Actuation Time 1450 gpm 26 sec SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating Fastests Post Accident Initiation of Fan Coolers 26 sec December 1976
~ k
LARGE BREAK Table 3 (continued)
CONTAINMENT DATA (Dry Containment)
STRUCTURAL HEAT SINKS Thickness (In)
Area (Ft2)
Steel Steel Concrete steel Steel Stainless Steel Concrete Stainless Steel Stainless Steel Stainless Steel Contrete 0.03 0,063 0.1 0.2 0.24
- 0. 2898, 24.0
- 0. 4896 0.6396 0.8904 1.256 1.56 2.0 2.75 5.5 9.0 0.44
- 2. 126 0.007 24.0 31,400 107,158 56,371 57,185 9931 136,000 23,667 6537 4915 27802 5307 668
. 1268,7 1277.4 260 '
14,392 768 3704 102,400 59,132
. 12 December 1976
TABLE 4 Reflood Mass and Energy Releases for limiting.
Case DECLG (DC=0.4) and 5% Stm.
Gen.
Tube. Plugging TIME (Sec)
TOTAL MASS FLOWRATE (LBm/Sec)
TOT~ ENERGY FLOWRATE (105 BTU/Sec) 46.943
- 49. 068 54.575 64.668 77.168 91.868 107.468 123.568 157.968 195.568 0.0 0.0
- 35. 89 99.81 101.9 107. 0 215.6;,.
260. 7 273. 1 279.8 0.0 0.0 0.4663 1.239 1.264 1.319 1.623
- 1. 707 1.638 1.545 13 December 1976
TABLE 5 Broken Loop Accumulator Flow to Containment For Limiting Case DECLG (CD=0.4) 5% Stm Gen.
Tube Plugging Time (Sec)
Mass Flowrate
( LBm/Sec) 0.0 0.02 2.00 4,00 6.00 8.00 10.0 15.0 20.0 25.0 30.0 35.0 38.9 0.0 2723.5 2276.0 1994.9 1793.1 1645.5 1526.4 1302.5 1137.8 1034.6 954.2 887.0 842.6 December 1976
~ 1 A
.9
.8
.5 IO 2
5 IO'IME (SECONDS}
lp~
2 IP3 Figure la.
Fluid Quality DECLG (Cg = 1.0)
I
.95 I
Pu
.90 CD
~ 75
.70 IO 2
IO 2
5 IO 2
TIME (SECONDS)
IO 2
Figure 1b.
Fluid Quality OECLG (CD = 0.6)
I
.7
.6 lpo 5
lpi TIME (SECOHDS)
!02 2
I03
- Figure 1c.
Fluid Quality DECLG (CD = 0.4)
l l
0
I.2 I.P
.7
.6 IO 2
IO 2
5 IO 2
TIME (SECOIIDS)
IO 2
Figure 1d.
Fluid Quality DECLG (CD = 0.4)
50 0
-50 I
-l00 I
)
ce l50
-200
-250lo-'O 2
5 lOi 2
TIME (SECONDS)
I02 2
Figure 2a.
Mass Velocity DECLG (Cp = 1.0)
lpp 50 cn 0
I
-50 CD ED O
g
-IOO 2*
- l50
-200lp-'p 2
5 IO' TIME (SECONDS)
IP 2
IP3 Figure 2b.
Mass Velocity DECLG (CD = 0.6)
30 IO I
0
- l0
-20
!0' Io'.
5 Io'IME (SECONDS) t Ip 2
Figure 2c.
Mass Velocity DECLG (Cp = 0.4)
k 4
'40 30 20 I
-IO
-20 IO 2
5 IO 2
5 IO 2
-TIME (SECONDS) 10 2
5 IO Figure 2d.
Mass Velocity -. DECLG (Cg
= 0.4)
l03 O
I 2
I l02 I
UJ CD U
5 5.75'0 0
l00 200 TIME (SECONDS) 300 400-500 Figure 3a.
Heat Transfer Coefficient DECLG (CD = 1.0)
CO I
CO
S 0'
lO~
5 OI5 2
I CVI lO~
5 Ug 2
los I
I 5.75'.25'o'00 200 300 TIME (SECONDS)
Figure 3b.
Heat Transfer Coefficient DECLG (Cp = 0.6)
/
C)
COI CO
IO~
I 5
ED 2
IO'-
I 5
IO' IOO TIME (SECONDS) 500 Figure 3c.
Heat Transfer Coefficient DECLG (Cp = 0.4)
IO'OO 200 TIME (SECONDS) 300 tIOO 500 Figure 3d.
Heat Transfer Coefficient -- DECLG (Cg
= 0.4)
f0. 749-l3 C)
CV II Cl O
lO CD I
G9 Ch ED UJ C/0 UJ I-O 0
Q)
L CL 0) 0O (visd) sznssszd
2500 2000
~
1500 1000 500 0
10 TIME (SECONDS) 15 20 25 Figure 4b.
Core Pressure DECLG (Cp = 0.6)
2500 2000 I500 0
0 IO 20 TIME (SECONDS) 30 40 50 Figure 4c.
Core Pressure DECLG (Cp = 0.4)
O.,
2500 2000 1500 ul 1000 0
0 10 20 TIME (SECONDS) 30 40 50 Figure 4d.
Core Pressure DECLG (Cp = 0.4)
s x i04 6 X lO" e X lO" ED 2X l04 0
-2X 10" 0
lo TIME (SECONDS) l5 20 25 Figure Sa.
Break Flow Rate DECLG (CD = 1.0I
IX IOS 8 X I04 6 x i04 e x i04 2X 10" 0
-2X IO'i 0
lo TIME (SECONDS) l5 20 25 Figure 5b.
Break Flow Rate DECLG (CD = 0.6)
II 0
i.i x ios S X IO" 7X IO 5X IO" hC 3 X i04 I x io" 0
-i x i04 0
'I0 20 TIME (SECOHDS) 30 50 Figure 5c.
Break Flow Rate DECLG (CD = 0.4)
9X IO4 7X IO 5
X I04 hC 3
X I04 I
X I04 0
-I X IO 0
IO 20 TIME (SECONDS) 30 50 Figure 5d.
Break Flow Rate DECLG (Cg = 0.4)
70 50 25 0
o
-25
-50
-70 0
IO TIME (SECONDS)
I5 25 Figure 6a.
Core Pressure Drop DECLG (CD = 1.0)
70 50 25 0
o
-25
-50
-70 0
10 TIME (SECONDS) l5 20 25 figure 6b.
Core Pressure Drop DECLG (CD = 0.6)
I 0'
70 50 25 Cl CO5 UJ 0
-50
-70 0
IO 20 TIME (SECONDS) 30 QO 50 Figure 6c.
Core Pressure Drop DECLG (CD = 0.4)
0"
70 50 25 Cl UJ 0
UJo -25 C)
-50
-70 0
IO 20 TIME (SECONDS) 30 40 50 Figure 6d.
Core Pressure Drop DECLG {CD = 0.4)
4
2500 o
2000 C5 CO g
I500 i5 g
I000 500 6'.
75' 0
100 TIME (SECONDS)
Figure 7a.
Peak Clad Temperature DECLG (Cp = 1.0)
2500 o
2000 Ch CO g
l500 au I 000 500 6.
25'.75' 0
l00 200 TIME (SECONDS) 500 Figure 7b.
Peak CIad Temperature DECLG (Cp = 0.6)
0
2500 o
2000 Ch ED I
g l500 I
5 l000 Ch 500 0
0 l00 TIME (SECONDS) 300
%00 Figure 7c.
Peak Clad Temperature DECLG (Cp = 0.4)
0 4
2500 L
2000 C5 CD I 500 5
IOOO CS 500 IOO 200 TIME (SECONDS) 400 Figure 7d.
Peak Clad Temperature DECLG (Cp = 0.4)
l
l750 l500 I250 l000 750 500 250 0
l00 TIME (SECONDS) 300 Figure Sa.
Fluid Temperature DECLG (Cp = 1.0)
C)
COI CO
2000 l750 l500 I250 l000 h
750 500 250 IOO 200 TIME (SECONDS) 300 400 500 Figure Sb.
Fluid Temperature DECLG (CD = 0.6)
I750 I 500 I000 I~
750 250 200
'00 TIME (SECONDS)
Figure 8c.
Fluid Temperature DECl G (CD = 0.4)
2000 I750 I500 I250 I000 750 500 250 200 TIME (SECONDS)
%00 Figure Sd.
Fluid Temperature DECLG (CD = 0.4)
l0.709-33 CD CV II CI O
U O
C5 I
E0 0
Ch CD UJ UJK I
P0O CD CV CD LA I
7000 5000 2500 0
-2500 TOP BOTTOH
-5000
-7000 IO TIME (SECONDS)
I5 20 Figure 9b.
Core Flow Top and Bottom DECLG (CD = 0.6)
7000 5000 2500 BOTTOM 0
C)
-2500
-7000 l0 20 TIME (SECONDS) 30 50 Figure 9c.
Core Flow Top and Bottom DECLG (CD = 0.4)
7000 5000 2500 TOP 0
ED IM -2500 BOTTOM
-7000 0
IO 20 TIME (SECONDS) 30 50 Figure 9d.
Core Flow Top and Bottom DECLG (CD = 0.4)
20.0
- 2. 00 l7. 5 OOWNCOHER LEVEL I.75 l5.0 l.50 l0.0 7.5 ViN I. 25 I.00 I
75 3 5.0 CORE LEVEL
.50 2.5
.25 40 50 75 I00 l25 l50 TIME (SECONDS) l75 200 225 Figure 10a.
Reflood Transient DECLG (Cp = 1.0)
Downcomer and Core Water Levels
20.0 2.00
- 17. 5 OOWNCOHER LEVEL l.75 le 0 I.50 Io.o 7.5
$ N I. 25 g I.00 I
.75 8 5.0 CORE LEVEL
.50 2.5
.25 50 75 100 125 I 50 TIME (SECONDS)
I75 225 Figure 10b.
Reflood Transient DECLG {Cp = 0.6)
Downcomer and Core Water Levels
h C
I'
20.0 2.00
- 17. 5 OOWXCOHER LEVEL 1.75 15.0 1.50
- 12. 5 10.0 7.5 ViX I. 25 1.00 I
75 cD 5 ~ 0 CORE LEVEL
.50 2.5
.25 50 75 100 125 150 T IME (SECONDS) 175 200 225 Figure 10c.
Reflood Transient DECLG (Cp = 0.4)
Downcomer and Core Water Levels
q ~
~
20.0 2.00 I7.5 OOWNCOMER LEVEL I.75 I5.0 I. 50 l2.5 l
10.0 7 5 l.25 ~
l.00 I-Cl
.75 o 5.0 CORE LEVEL
.50 2.5
.25 50 75 l00 I25 I50 TIME {SECONDS)
I75 200 225 Figure 10d.
Reflood Transient DECLG (CD = 0.4)
Downcomer and Core Water Levels
V
5000 co
%000 3000 2000 l000 0
0 IO TIME (SECONDS) 20 25 Figure 11a.
Accumulator Flow (Blowdown) DECLG (Cg = 1.0)
r 4
6000 5000
%000 3000 2000 I000 0
0 IO TIME (SECONOS) l5 20 25 Figure 11b.
Accumulator Flow (Slowdown) DECLG (CD = 0.6)
6000 5000 g
%000 CD 3000 8
2000 l000 0
0 IO 20 TIME (SECONDS) 30 50 Figure 11c.
Accumulator Flow (Blowdown) DECLG (Cg = 0.4)
P
~ >>
U
6000 5000 4000 ED 3000 I
8 2000 I000 0
0 lp 20 30 TIME (SECONDS) 40 50 Figure 11d.
Accumulator Flow (Blowdown) DECLG (CD = 0.4)
0',
l2 10 0
0 80 I20 I60 200 TIME (SECONDS) 280 320 Figure 12a.
Pumped ECCS Flow (Reflood) DECLG (CO = 1.0)
1
l2 IO 80 I20 I60 200 TIME (SECONDS) 280 320 360 Figure 12b.
Pumped ECCS Flow (Reflood) DECLG (Co = 0.6)
IO 6
80 I20 I60 200 TIME (SECONDS) 280 320 360 Figure 12c.
Pumped ECCS Flow (Reflood) DECLG (Cp = 0.4)
I
l0 80 I 20 I60 200 TIME (SECONDS) 2%0 280 320 360 Figure 12d.
Pumped ECCS Flow (Reflood) DECLG (CD = 0.4)
35 30 25 20 l-IS I
TIME (SECONDS) 280 320 360 Figure 13a.
Containment Pressure DECLG (Cp = 1.0j
4 f
35 30 25 I
lxJ I5 I
CO CD
. IO 80 I20 I 60 200 2%
TIME (SECONDS) 280 320 360 Figure 13b.
Containment Pressure DECLG (Cp = 0.6)
CO
'4 COI col CD
'1 P
p
30 25 a
20 0'0 80 I20 I 60 200 240 TIME (SECONDS) 280 320 360 400 Figure 13c.
Containment Pressure DECLG (Cp = 0.4)
30 25 C9 2
20 LLI l5 I
IO 5
CD 80 I20 I60 200 2tIO TIME (SECONDS) 280 320 360 Figure 13d.
Containment Pressure pECLG (Cp = 04)
.2 0
0 IO TIME (SECONDS) l5 20 25 Figure 14a.
Core Power Transient DECLG (Cp = 1.0j
.8
.6
.2 0'
IO TIME (SECONDS) l5 20 25 Figure 14b.
Core Power Transient pECLG (Cp = 0.6)
.8
.2 0
0 IO 20 TIME (SECONDS) 30
"%0 50 Figure 14c.
Core Power Transient DECLG (Cp = 0.4)
~ 8 0
20 30 TIME (SECONDS)
%0 50 Figure 14d.
Core Power Transient DECLG (Cp = 0.4)
<<C 1.5 X l07.
-.5 X l06 0
IO 20 30 TIME (SECONDS) 50 Figure 15.
Break Energy Released to Containment
( CD = 0.4)
q,
l0.7%9-53 I000 U0 I
CVI U-I I
CO I-UJ UUUJ UJ U
I-CO C/I UJ Ch ED I
UJ l
900 800 700 600 500 400 300 200 l00 l00 TIME (SECONDS) 200 300 Figure 16.
Containment Wall Condensing Heat Transfer Coefficient (CD = 0.4)
laMCM