L-76-419, Letter Regarding ECCS Reevaluation

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

U.S. NUCLEAR REGULATORY COIAMISS ION DOCK 8ER hlRC FORM 195 I2~ 76 I -2 251 NRC DISTRIBUTION soR PART EO DOCKLI P MA I'ERIAL TO: FROM: DATE OF DOCUMENT Florida Power & Light Company 12/9/76 Hr. Victor Stello Hiami, Florida DATE RECEIVED Hr. Robert E. Uhrig 12/9/76

%LETTER DNOTORIZED PROP INPUT FORM NUMSER OF COPIES RECEIVED JHl ORIGINAL gUNC LASSIF IE D Three signed Qcopv 35 copies encl recvd.

DESCRIPTION ENCLOSURE t

L tr. w/a tached erra ta shee t.... re our 6/17/76 and 12/3/76 orders...trans the following: Concerns re-evaluation of ECCS cooling performance calculated in accordance with SOMOT REMOVE an approved Nestinghouse Evaluation Model, with appropriate corrections for upper head water temperature.

ACKNOWLEDGED (3-P) (6O-P) .

PLANT NAME:

Turkey POIInt Units 3 & 4 SAFETY FOR ACTION/INFORMATION 12 10 76 ASSXGNED AD:

Lear RO EC MANA Elliott PROJECT MANAGER'IC L C ASST arrz.s ASST INTERNAL DISTRIBUTION REG FILE SYSTEMS SAFETY PLA1AT SYSTEMS SITE SA E HEINEHAN TEDESCO I &E SCHROEDER N 0 A OELD GOSSICK & STAFF ENGINEERING IPPOLXTO MIPC ERNST CASE KNXGHT HANAUER HARLESS SIHWEIL PAWL CK OPERATING REACTORS STELLO SPANGLER gpss SITE TECH PROJECT MANAGEMENT REACTOR SAFE OPERATING TECH GAMHILL BOYD ROSS EXSENHUT STEPP PE COLLINS NOVAK HULMAN HOUSTON ROSZTOCZY PETERSON CHECK B .E SITE ANALYSXS MELTZ VOLLHER HELTEMES AT & I BUNCH SKOVHOLT SALTZMAN J ~ COLLINS RUTBERG KREGER EXTERNAL DISTRIBUTION, CONTROL NUMBER LPDR ~ Miami, Fla. NAT LAB'EG B 0 K TXC: VOGIE ULR KSON OR NSIC: LA PDR j+

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~~+ FLORIOA POWER & LIGHT COMPANY y'~c @@4 December 9, 1976 L-76-419 Office of Nuclear Reactor Regulation 0 Attention: Mr. Victor Stello, Director Division of Operating Reactors <-'~'~o~

U. S. Nuclear Regulatory Commission +

Washington, D. C. 20555 +e pegutatory Docket I Iie

Dear Mr. Stello:

Re: Turkey Point Units 3 6 4 Docket Nos. 50-250 and 50-251 ECCS Re-evaluation 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 Electric Corporation is now revising their calcula-Mwt.'estinghouse 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".

e ~ <a MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT)

The analysis specified by 10CFR50.46 Cl] Acceptance Criteria for Emer-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.

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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

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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 These figures show the hot spot clad temperature transient through 9d 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 These figures show the Emergency Core Cooling System through 12d 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 The containment pressure transient is also provided.

a,b,c,d Figures 14 These figures show the core power transient.

a,b,c,d Figures 15 These figures show the break energy released to the a,b,c,d 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.

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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 +DECL ~DECL o'ctcDECLG (CD-1.0) (CD=0.6) (CD=0.4) (CD=0.4)

(Sec) (Sec) (Sec) (Sec)

START 0.0 0.0 0.0 0.0 Reactor Trip Signal 0.573 0.583 0,595 0.595 S. I'ignal 0.44 0.55 0.67 0.67 Acc. Injection 10. 3 12 ' 16.6 16.3 End of Bypass 20.96 23. 81 27. 89 27.66 End of Blowdown 21.18 23.96 28 04

~ 27.84 Bottom of Core Recovery 40.37 42.78 46.94 46.78 Acc. Empty 55.09 57.71 61. 52 61. 19 Pump Injection 25.44 25.55 25.67 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 >>DECL '<DECL <<DF.CL >>>>DECLG (CD-1.0) (CD+0.6) (CD=0.4 (CD=0.4 Results Peak Clad Temp. 'F 1767 1927 2162 2198 Peak Clad Location Ft. 6.25 6.5 6.0 6.0 Local Zr/H20 Rxn(max)% 2.009 3.627 10.732 12.310 Local Zr/H20 Location Ft. 6,0 6.0 6.0 6' Total Zr/H20 Rxn % L0.3 L0.3 LO ~ 3 L0.3 Hot Rod Burst Time Sec. 45.24 31.0 24.1 22,9 Hog Rod.Burst Location Ft. 6 0

~ 6.0 6.0 6.0 Calculation Core Power'wt 102% of 2192 Peak Linear Power kw/ft 102% of Peaking Factor 2.25 Accumulator Mater Volume (ft3) ~875 ( ar accumulator)

Fuel region + cucle analyzed Cycle Region UNITS 3 and 4

• Contains 5% Stm. Gen. Tube Pilugging

>>>> Contains 10% Stm Gen. Tube Plugging December 1976 10

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TABLE 3 LARGE BREAK CONTAINNFNT DATA (DRY CONTAINMENT)

NET FREE VOLlNi E 1.55xl0 Ft3 INITIAL CONDITIONS Pressure 14.7 psia 900&

Temperature RMST Temperature 39'F Service Hater Temperature 63'F Outside Temperature 39'F SPRAY SYSTEM Number of Pumps Operating Runout Flow Rate 1450 gpm Actuation Time 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 0.03 31,400 0,063 107,158 0.1 56,371 0.2 57,185 0.24 9931 Steel 0. 2898, 136,000 Concrete 24.0 steel 0. 4896 23,667 0.6396 6537 0.8904 4915 1.256 27802 1.56 5307 2.0 668 2.75 . 1268,7 5.5 1277.4 Steel 9.0 260 '

Stainless Steel 14,392 Concrete Stainless Steel 0.44 768 Stainless Steel 2. 126 3704 Stainless Steel 0.007 102,400 Contrete 24.0 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 TOTAL MASS FLOWRATE TOT~ ENERGY FLOWRATE (Sec) (LBm/Sec) (105 BTU/Sec) 46.943 0.0 0.0

49. 068 0.0 0.0 54.575 35. 89 0.4663 64.668 99.81 1.239 77.168 101.9 1.264 91.868 107. 0 1.319 107.468 215.6;,. 1.623 123.568 260. 7 1. 707 157.968 273. 1 1.638 195.568 279.8 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 0.0 0.02 2723.5 2.00 2276.0 4,00 1994.9 6.00 1793.1 8.00 1645.5 10.0 1526.4 15.0 1302.5 20.0 1137.8 25.0 1034.6 30.0 954.2 35.0 887.0 38.9 842.6 December 1976

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Figure 5d. Break Flow Rate DECLG (Cg = 0.4)

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Figure 6a. Core Pressure Drop DECLG (CD = 1.0)

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Figure 6c. Core Pressure Drop DECLG (CD = 0.4)

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Figure 6d. Core Pressure Drop DECLG {CD = 0.4)

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Figure 7b. Peak CIad Temperature DECLG (Cp = 0.6)

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Figure 7c. Peak Clad Temperature DECLG (Cp = 0.4)

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Figure 7d. Peak Clad Temperature DECLG (Cp = 0.4)

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Figure 8c. Fluid Temperature DECl G (CD = 0.4)

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Figure Sd. Fluid Temperature DECLG (CD = 0.4)

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Figure 9c. Core Flow Top and Bottom DECLG (CD = 0.4)

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-7000 0 IO 20 30 50 TIME (SECONDS)

Figure 9d. Core Flow Top and Bottom DECLG (CD = 0.4)

20.0 2. 00 l7. 5 I.75 OOWNCOHER LEVEL l5.0 l.50 ViN I. 25 l0.0 I .00 I

7.5 75 3 CORE LEVEL 5.0 .50 2.5 .25 40 50 75 I00 l25 l50 l75 200 225 TIME (SECONDS)

Figure 10a. Ref lood Transient DECLG (Cp = 1.0)

Downcomer and Core Water Levels

20.0 2.00

17. 5 l.75 OOWNCOHER LEVEL le 0 I.50

$ N I. 25 g Io.o I.00 I 7.5 .75 8 CORE LEVEL 5.0 .50 2.5 .25 50 75 100 125 I 50 I75 225 TIME (SECONDS)

Figure 10b. Ref lood Transient DECLG {Cp = 0.6)

Downcomer and Core Water Levels

h C

I'

20.0 2.00

17. 5 1.75 OOWXCOHER LEVEL 15.0 1.50

12. 5 I. 25 ViX 10.0 1.00 I

cD 7.5 75 CORE LEVEL 5~0 .50 2.5 .25 50 75 100 125 150 175 200 225 T IME (SECONDS)

Figure 10c. Ref lood Transient DECLG (Cp = 0.4)

Downcomer and Core Water Levels

q~

• ~

20.0 2.00 I7.5 I.75 OOWNCOMER LEVEL I 5.0 I. 50 l.25 ~

l l2.5 10.0 l.00 I-Cl 7 5 .75 o CORE LEVEL .50 5.0 2.5 .25 50 75 l00 I25 . I50 I75 200 225 TIME {SECONDS)

Figure 10d. Ref lood Transient DECLG (CD = 0.4)

Downcomer and Core Water Levels

V 5000 co %000 3000 2000 l000 0

0 IO 20 25 TIME (SECONDS)

Figure 11a. Accumulator Flow (Blowdown) DECLG (Cg = 1.0)

r 4 6000 5000

%000 3000 2000 I 000 0

0 IO l5 20 25 TIME (SECONOS)

Figure 11b. Accumulator Flow (Slowdown) DECLG (CD = 0.6)

6000 5000 g %000 CD 3000 8 2000 l000 0

0 IO 20 30 50 TIME (SECONDS)

Figure 11c. Accumulator Flow (Blowdown) DECLG (Cg = 0.4)

P

~ >>

U

6000 5000 4000 ED 3000 I

8 2000 I 000 0

0 lp 20 30 40 50 TIME (SECONDS)

Figure 11d. Accumulator Flow (Blowdown) DECLG (CD = 0.4)

0',

l2 10 0

0 80 I20 I60 200 280 320 TIME (SECONDS)

Figure 12a. Pumped ECCS Flow (Ref lood) DECLG (CO = 1.0)

1 l2 IO 80 I20 I60 200 280 320 360 TIME (SECONDS)

Figure 12b. Pumped ECCS Flow (Ref lood) DECLG (Co = 0.6)

IO 6

80 I20 I60 200 280 320 360 TIME (SECONDS)

Figure 12c. Pumped ECCS Flow (Ref lood) DECLG (Cp = 0.4)

I

l0 80 I 20 I60 200 2%0 280 320 360 TIME (SECONDS)

Figure 12d. Pumped ECCS Flow (Ref lood) DECLG (CD = 0.4)

35 30 25 20 l-IS I

ED IO 80 120 I 60 200 2% 280 320 360 TIME (SECONDS)

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% 280 320 360 TIME (SECONDS)

CO

'4 Figure 13b. Containment Pressure DECLG (Cp = 0.6) CO I

col CD

'1 P p

30 25 a 20 0 '0 80 I20 I 60 200 TIME (SECONDS) 240 280 320 360 400 Figure 13c. Containment Pressure DECLG (Cp = 0.4)

30 25 C9 2 20 LLI l5 I

IO 5

CD 80 I20 I 60 200 2tIO 280 320 360 TIME (SECONDS)

Figure 13d. Containment Pressure pECLG (Cp = 04)

.2 0

0 IO l5 20 25 TIME (SECONDS)

Figure 14a. Core Power Transient DECLG (Cp = 1.0j

.8

.6

.2 0'

IO l5 20 25 TIME (SECONDS)

Figure 14b. Core Power Transient pECLG (Cp = 0.6)

.8

.2 0

0 IO 20 30 "%0 50 TIME (SECONDS)

Figure 14c. Core Power Transient DECLG (Cp = 0.4)

~ 8 0

20 30 %0 50 TIME (SECONDS)

Figure 14d. Core Power Transient DECLG (Cp = 0.4)

LI.5 X I07 3.5 X 107 2.5 X IO

<<C 1.5 X l07.

~ SX IO'

-.5 X l06 0 IO 20 30 50 TIME (SECONDS)

Figure 15. Break Energy Released to Containment ( CD = 0.4)

q, l0.7%9-53 I 000 900 U

0 I

CV I

U-I 800 I

CO I- 700 UJ U

U UJ 600 UJ U

I- 500 CO C/I 400 UJ Ch ED 300 I

UJ 200 l

l00 l00 200 300 TIME (SECONDS)

Figure 16. Containment Wall Condensing Heat Transfer Coefficient (CD = 0.4)

laMC M