ML19339B278

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Draft Proposed Tech Specs Change 117,describing Safety Injection Sys Protection Against LOCA
ML19339B278
Person / Time
Site: Yankee Rowe
Issue date: 08/02/1979
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YANKEE ATOMIC ELECTRIC CO.
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ML19339B279 List:
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NUDOCS 8011060658
Download: ML19339B278 (33)
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{{#Wiki_filter:w.I v m g gg h 41 1 8/2 /74 \\ r 'h' M 411 LOSS-OF-COOLANT ACCIDENT l GENERAL t A loss-of-coolant accident is defined as a rupture of the main coolant system piping which results in interruption of the normal mechanism for removing heat from the reactor core. The Yankee Nuclear Power Station is equipped with a safety injection system (see Section 212) which assures adequate protection for the core in the unlikely event of a loss-of-coolant accident by providing injection flow of borated water to cool the core and maintain a shutdown condition. Should a major break occur, depressurization of the main coolant system results in a pressure decrease in the pressurizer. A reactor trip signal occurs when the pressurizer or main coolant system low pressure trip setpoint is reached. A safety injection system signal is actuated when the appropriate setpoint is reached. Saf ety injection i system actuation is also provided by a high containment pressure signal. These countermeasures will limit the consequences of the accident in l two ways: (1) Reactor trip and borated water injection complement void formation in causing the rapid reduction of power to a residual level corresponding to delayed fission and fission product decay heat. p) (2) Injection of borated water provides heat transfer from core and ( prevents excessive clad temperature. The safety injection system consists of eafety injection storage tank, three low pressure injection pun.n, ec 2 accumulator with its nitrogen supply, three high pressure inj.:ction pumps and associated pipes, valves and instrumentation. Safety injection initiation is automa low main coolant or low pressurizer pressure generates a safet- _ cion and alarm signal (SIAS). The low pressure safety injection (LPS1/,.mps take suction from the safety injection storage tank and discharge to the LPS1 header and/or to the high pres'sure safety injection (HPSI) suction header. The HPSI pumps take suction from the discharge of the LPSI pumps and discharge to the HPSI header. Safety injection flow enters the main coolant system loops through four lines branching from the safety injection header in the vapor container. Flowmeters are provided in each of these lines. Excessive I flow in any one of the four lines will indicate to the operator the l possibility of a break in that line. The operator could then isolate i that line with either one of the redundant remotely operated branch block valves and continue safety injection through the other three lines. l The accumulator is a passive component which is pressurized to an assumed ,r j minimum of 420 psia with nitrogen from the high pressure N storage bottles. 2 During normal operation, the accumulator is separated from main coolant system 1 8011000 h $ @ t

Tk t 411:02 8/2 /74 DRAFT pressure by several check valves. Flow from the accumulator starts auto-matically when the main coolant system pressure falls below the pressure of the cccumulator. A constant pressure is maintained in the accumulator by the gas pressure regulating valves. After sufficient volume has flowed from the accumulator, a level control signal vents the nitrogen through the pilot operated relief valves. All active components of the saf ety injection system are redundant. The system will perform as required after the failure of any single active component and a loss of normal AC power. Safety injection system power is furnished from three 480 volt emergency busses. Each bus is powered by a 500 KVA diesel generator if normal AC power is lost. Each bus powers one LPSI pump and one HPSI pump. Where a valve may be required to close during the course of an acc id en t, there are two valves in series so that flow will be stopped if only one valve operates. Stailarly, where a valve is required to open, there are two valves in parallel and each is sized for 100 percent flow (for example, the accumulator vent valves or the nitrogen regulators). In order to evaluate the loss-of-coolant accident, a range of break sizes from small leaks up to a complete double-ended severance of a g', main coolant pipe has been considered. t i a METHOD OF ANALYSIS AND RESULTS The analysis specified by 10CFR50.46 , " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors", is presented in this section. The analytical techniques used are in compliance with Appendix K of 10CFR50 and are deggyibed in the topical report, " Westinghouse ECCS Evaluation Model - Summary" The exceptions to the assumptions describ a in Reference 6 are: (1) The rated power level is increased by 3 percent to allow for instrumentation errors. inthefuelandcladdingis0.87,basedoncalculations{ydeposited The fraction of the locally generated gamma energy that (2) for the Yankee core (which utilizes cruciform control rods). (3) The worst single failure is determined to be the failure of a diesel to start following the postulated loss of offsite power. The method of analysis used for the larger size breaks is somewhot different than the method used for the smaller size breaks. Small Break Analysis l Ruptures of small cross section will cause expulsion of the coolant at a rate which can be accommodated by the charging pumps which would J

[ .s '411*.03 8/2./74 DRAFT _j O maintain an operational water level in the r essurizer permitting the operator to execute.an orderly shutdown. The maximum break size for which the normal ukeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the main coolant system through the postulated break against the charging system makeup flow at maxinum operating main coolant system pressure, i.e., 2090 psia. 'A makeup flow rate from the charging. system is typically adequate to sustain pressurizer level at 2090 psia for a break through a 3/8 inch diameter hole. This break results in a loss of approximately ~ 7.2 lb/sec. Should'a larger break' occur, the inability of the charging system to maintain pressurizer level results in a depressurization of the main coolant system. Prior to the break, the plant is at equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. During blowdown, heat from decay, hot internals and the vessel continues to be transferred to the main coolant system. The heat transfer between the main coolant system and the secondary system may be in either direction depending on the relative temperatures. In the case of continued. heat addition to the secondary, steam pressure increases and steam relief through the dump or safety valves may occur. Make-up to the secondary side is provided by the auxiliary feedwater system.- The secondary flow aids in the reduction of main coolant system pressure. .When the main coolant system depressurizes to 420_ psia, the accumulator begins to inject water into the main coolant loops. The main coolant J pumps are assumed to be tripped at the initiation of the accident and the effects of pump coastdown are included in the blowdown analyses. For small breaks (less than 1.0 ft ) the WFLASH( } digital computer code is employed to calculate the transient depressurization of the main coolant system as well as to describe the mass and enthalpy of flow through the break. TheWFLASHprogramusedintheanalysisofge small break loss of coolant accident is an extension of tM FLASH-4 code developed at the Westinghouse Bettis Atomic Power Laboratory..The ~ WFLASH program permits.a detailed spatial representation of the main coolant system. The main coolant system is nodalized into volumes interconnected by ) flowpaths. The broken loop is modeled explicitly with the intact loops lumped together. The transient behavior of the system is determined from the governing conservation equations of mass, energy and momentum applied throughout the system. The use of WFLASH in the analysis involves, among other things, the representation of the reactor core as a heated control volume with the associated bubble rise model to permit a transient mixture height calculation. The multi-node capability of the program enables an explicit and detailed spatial representation of various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant transient. b

4 1 411".04 8/2 /74 DRAFT o( Safety injection flow rate to the main coolant system as a function of the system pressure is used as part of the input. The safety injection (SI) system was assumed to be delivering full flow to the main coolant system 40 seconds after the generation of a safety injection signal. For these analyses, the safety injection delivery considers the injection flow which is depicted in Figure 411-1 as a function of main coolant system pressure. The 40 second delay includes time required for diesel startup and loading of the safety injection pumps onto the anergency busses. Minimum Emergency Core Cooling System capability ard operability has been assumed in these analyses. Peak clad temperature analyses are performed with the LOCTA IV program which is described in Reference 8. The small breaks analyzed were 2, 3, and 4 inch diameter ruptures. The worst break size is a 4 inch diameter break. The depressurization transient for this break is shown in Figure 411-2. The extent to which the core is uncovered is shown in Figure 411-3. During the early part of the small break transient, the effect of the break flow is not strong enough to overcome the flow maintained by the main coolant pumps through the core as they are coasting down following reactor trip. Therefore, upward flow th.ough the core is maintained. The resultant heat transfer cools the fuel rod and clad 7s (_j to very near the coolant temperatures as long as the core remains covered by a two phase mixture. The maximum hot spot clad temperature calculated during the transient is 1265'F including the effects of fuel densification as described in Reference 4. The peak clad temperature transient is shown in Figure 411-4 for the worst break size; i.e., the break with the highest peak clad temperature. The steam flow rate for the worst break is shown in Figure 411-5. When the mixture level drops below the top of the core, the steam flow computed in WFLASH is used to determine the cooling to the upper portion of the core. The rod film coefficients for this phase of the transient are given in Figure 411-6. The hot spot fluid temperature for the worst break is shown in Figure 411-7 and the normalized core power transient following the accident (relative to reactor scram time) is shown in Figure 411-8. The reactor shutdown time is equal to the reactor trip signal time pluse 3.2 sec for rod insertion and delay. During this rod insertion period the reactor is conservatively assumed to operate at rated power. Additional break sizes were analyzed. Figures 411-9A and 411-9B present the main coolant system pressure transient for the 3 and 2 inch breaks respectively and Figures 411-10A and 411-10B present the volume history (mixture height) plots for both breaks. The peak clad tempera-tures for both cases are less than the peak clad temperature of the q \\

e \\ 411.05 8/2./74 O DRAFT 4 inch break. The peak clad temperatures for the 3 inch break is given in Figure 411-11. The peak clad temperature for the 2 inch break is not shown since the core is not uncovered during the transient. 1 The analyses presented demonstrate that the Emergency Core Cooling System provides sufficient core flooding to keep the calculated peak clad temperatures below the required limits of 10CFR50.46. Hence, adequate protection is afforded by the Emergency Core Cooling System in the event of a small break loss of coolant accident. Large Break Analysis At the beginning of the blowdown phase of a major break, the entire main coolant system contains subcooled liquid which transfers heat from the core by forced convection uith some fully developed nucleate boiling. After the break develops, the time to departure f g nucleate boiling is calculated, consistent with Appendix K of 10CFR50 Thereafter, the core heat transfer is unstable, with both nucleate boiling and film boiling occurring. As the core becomes uncovered both turbulent and laminar forced convection and radiation are considered as core heat transfer mechanisms. When the main coolant system pressure falls below 420 psia the accumulator begins to inject borated water. The conservative assumption is made that accuuulator water injected bypasses the core and goes out (') through the break until the termination of bypass. This conservatism is V again consistent with Appenxix K of 10CFR50. The Emergency Core Cooling System, even when operating during the injection mode with most severe single active failure meets the l Acceptance Criteria' The descript of the various aspects of the LOCA analysis is given in WCAP-8339 Thio document describes the major phenomena modeled, the interfaces among the computer codes and features of the codes which maintain compliance with the Acceptance Crigr g The individual codes ) are described in detail in separate reports The containment parameters used in the containment pressure code for determination of ECCS backpressure are given in Table 411-2. Table 411-1 presents the peak clad temperatures and hot spot metal reaction for a range of break sizes. The analysis of the loss of coolant accident is performed at 103 percent of rated power. The peak linear power, and core power used in the analyses are also given in Table 411-1, along with the equivalent core peaking factor at the license application power level. Since there is a margin between 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 obtaincJ by using the peak linear power density expected during operation. OV t

e 4 411.06 8/2 /74 f) DRAFT Figures 411-12 through 411-24 present the transient values of the principle parameters for the break sizes analyzed. These figures present the following information: 411-12 Value of fluid quality at the clad burst location and at the hot spot. 411-13 Mass velocity at the clad burst location and at the hot spot. 411-14 Film heat transfer coefficient calculated by LOCTA IV at the clad burst location and at the hot spot. 411-15 Pressure calculated in the core. 411-16 Break flowrate. For guillotine breaks this is the sum of the flows out both ends of the break. 411-17 Core pressure drop from the lower plenum, near the core, to the upper plenum at the core outlet. 411-18 Clad temperature transient at the hot spot and at the burst location. 411-19 Fluid *'operature at the hot spot and turst location. r^% 4 i \\> 411-20 Cere flowrate (top and bottom). 411-21 Core reflood transient. 411-22 Accumulator flow. Established in the refill-reflood calculations and is the sum of that injected in the intact cold legs. 411-23 Pumped ECCS flow. 411-24 Containment pressure transient. The clad temperature analysis is based on a total peaking factor ) of 2.96. The hot spot metal water reaction reached is 8.6 percent, which is well below the embrittlement limit of 17 percent, as required by 10CFR50.46. In addition, the total core metal-water reaction is less than 0.3 percent for all breaks as compared with the 1 percent criterion of 10CFR50.46. The results of several sensitivity studies are reported in Reference 11. These results are for conditions which are not limiting in nature and hence are reported on a generic basis. CONCLUSIONS C) For breaks up to and including the double ended severance of a ( reactor coolant pipe:

-n. a 411.07 8/2 /74 p i -(1).The calculated peak fuel element clad temperature provides margin to the requirement of 2200*F. I 1 (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. (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 radioactivity remaining in the core. Therefore, the Emergency Core Cooling System will meet the Acceptance. i Criteria as presented in 10CFR50.46. REFERENCES. (1) J.' N. Hamawi, "lankee Rowe Core XI Decay Heat Redistribution Factor During Shutdown Conditions", YAFC-1071, June 1974. (2) V. J. Esposito, K. Kesavan, B. A. Maul, WFLASH - A FORTRAN IV Computer Program for Simulation of Transients in a Multi-Loop PWR, WCAP-8261 Rev. 1, July 1974. j (3) Porsching, T. A., Murphy, J. H., Redfield, J.'A., and Davis, V. C., " FLASH-4: A Fully Implicit FORTRAN-IV Program for the Digital Simulation of Transients in a Reactor Plant", WAPD-TM-84, } Bettis Atomic Power Laboratory (March, 1969). l (4) J. M. Hellman, Fuel Densification Experimental Results and Model for Reactor Application", WCAP-8219, October, 1973. 1 (5) " Acceptance Criteria for Emergency Core Cooling System for Light-Water ~ Cooled Nuclear Power Reactorst' 10CFR50.46 and Appendix K 4 j of 10CFR50. Federal Register, Volume 39, Numoer 3 January 4, 1974. (6) Bordelon, F. M., Massie, H. W., and Zordan, T. A., " Westinghouse ECCS Evaluation Model - Summary" WCAP-8399, July 1974. 1 (7) Bordelon, F. M., et al., " SATAN-V1 Program: Comprehensive Space-Time Dependent Analysio of Loss-of-Coolant", WCAP-8306, June 1974. (8) Bordelon, F. M., et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis", WCAP-8305, June 1974. 2 (9) Kelly, R. D., et al., " Calculational Model for Core Reflooding after a Loss-of-Coolant _ Accident (kjlEFLOOD Code)", WCAP-8171, June 1974. cO (10) Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (C0CO)",-WCAP-8326, June 1974 '(11) Buterbaugh, T. L., Johnson, W. J. and Kopelic, S. D., " Westinghouse _ECCS'- Plant Sensitivity Studies", WCAP-8356, July 1974.

>} ('~% mI w/ u, TABLE 411-1 IARGE BREAK ANALYSIS SUbMARY OF RESULTS* 2 Break Type DECLG** DECLG 0.6 x DECLS*** 1.0 ft 3 (CD = 1.0) (CD = 0.6) Peak Clad Temperature, F 2150 2175 2125 1850 Location of Peak, ft 3.8 3.8 3.8 4.1 Local blaximum Zr-Il 0 Reaction, % 7.1 8.6 6.5 2.6 Zr-fl 0 Reaction, ft 3.8 3.8 3.8 4.1 Location of blax. 2 Total Zr-Il 0 Reaction, % <0.3 <0.3 <0.3 <0.3 2 liot Rod Burst Time, sec 81.2 83.6 85.4 101.2 !!ot Rod Burst Location, it 3.8 3.8 3.8 3.8

  • Calculations are based on Reactor Power, bNt 618 Peak Linear Power, kw/ft 13.06 Peaking Factor at Rated 2.96 Power Accumulator Water Volume, 700 ft3 Fuel Cycle 11 omu Fuel Region B

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    • DECLG - Double-ended cold leg guillotine break.

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      • DECLS - Double-ended cold Icg slot break.

's 411.09 i 8/2 /74 DRAFT ("h TABLE 411-2 LARGE BREAK ANALYSIS CONTAINMENT DATA 3 Net Free Volume 903,000 ft Initial Conditions Pressure 14.7 psia c Temperature 90 p RWST Temperature 400F Service Water Temperature 33 F 0 Outside Temperature -30 F Structural lleat Sinks 2 Material (thickness) Area, Ft g-Concrete (60 in) 50,000 Carbon Steel (0.875 in) 49,000 Carbon Steel (0.055 in) 8,000 Stainless Steel (0.225 in) 10,800 Lj

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-LOSS OF COOLANT ACCIDENT FIGURE POWER STATION ECCS FLOW VS. PRESSURE 411-1A FOUR INCH DIAMETER BREAK U

~ 411:11 ( 8/2/74 DRAFT 1600 i 1400 i: 1200 'a a. 1000 4 E. 0 Ew' 800 U cs O 8 .5 600 "~ e l 400 r ~ ^200 0 i i e a 0 50 100 150 200 250 ECCS Flow, Ibm /sec om o g-6 O . m ;w m anu .\\ rrom YANKEU NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION ECCS FLOW VS. PRESSURE 411-1B 1WO AND TilREE INCil DIAMETER BREAKS

411'.12 8/2/74 DRAFT S ~ , c, m ,S t h o e h S S 8 S d" .P O S { m 4 l o S S S S S n S 3 cisd 'oanssold LOSS OF COOLANT ACCIDENT FIGURE YANKEE NUCLEAR POWER STATION MAIN COOLANT PRESSURE VS. TIME 411-2 FOUR INCH DIAMETER BREAK

411:13 hs 8/2/74 DRAFT 8 l l m I N N O 4 8 h 8 c. O ) i o g so O N m 7 da Y O / O A M ~ a b o o. m. o. m. o. e. o N 3 2003 '29ston aan2xm oaoa i 1

O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CORE COOLANT LEVEL VS. TIME 411-3 FOUR INCH DIAMETER BREAK j

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s 411:14 8/2/74 DRAFT O _g .. -....- - -. Q m g 8 ( _N n m E* m o .s g ~ O n 8 \\ ~ m 8 .= 8 8 8 8 8 8 8 m n o o m m d 'o.tnacJodmoJ, pcI3 O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE I POWER STATION PEAK CLAD TEMPERATURE VS. TIME 411-4 l FOUR INCH DIAMETER BREAK

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8 at ~ e o 8 8 8* e o o o i g a 2 ~. m. -cas/mqi 'o2 m old mea 2s O LOSS OF COOLANT ACCIDENT FIGURE YANKEE NUCLEAR POWER STATION CORE STEAM FLOWRATE VS. TIME 411-5 FOUR INCll DIAMETER BREAK s,, -,, ,---v.. w y, ~,, r v

411*.16 (^s 8/2/74 V DRAFT l 0 i 7 1[ll m i l T l .... y f n N p! L o* lj, i o l b m i n 'l l e g g i r4 o l J m O 5 l E-- V ^ / o q m O,- l n l = 26 A [ -.e g o m

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, a I i ,4 ! l l' ,l di m e o. o. o. ..o o uoT238Id *.romod Pozi185 ton 0103 3-9.A 4 b O fg 6 Ll u I o YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CORE POWER FRACTION VS. TIME 411-8 FOUR INCH DIAMETER BREAK

~. 411*.19 8/2/74 0 DRAFT 2500 2000 e 2 1500 R E \\ 1000 O \\ 500 0 0 100 200 300 400 500 600 700 Time, Seconds O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER "TATION MAIN COOLANT PRESSURE VS. TIME 411-9A THREE INCH DIAMETER BREAK

411'.20 V(^g 8/2/74 DRAFT 14.0 12.5 ( h Ahr g 10.0 y gy] W i I { TOP OF CORE - 7.6

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7.5 E i S* di 9 5.0 O 8 2.5 0 0 100 200 300 - 400 500 600 Time, Seconds YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CORE COOLANT LEVEL VS. TIME 411-10A THREE INCH DIAMETER BREAK

6 e 411".21 3 8/2/74 DRAFT 1000 900 u. 8 800 B E 700 O e f 3 600 G 500 400 100 140 180 220 Time, Seconds 1 O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION PEAK CLAD TEMPERATURE VS. TIME 411-11 THREE INCH DIAMETER BREAK

411'.22 8/2 /74 DRAFT 1 2500 2000 4 .5 J 1500 E Bc. i 1000 O 500 n 0 0 250 500 750 1000 1250 Time, Seconds i s YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION MAIN COOLANT PRESSURE VS TIME 411-9B TWO INCH DIAMETER BREAK

l 411.23 8/2/74 DRAFT i 14.0 12.5 ) g/M 10.0 i 3 TOP OF CORE - 7.6' o* 7.5 8 B i .5x 5.0 e O u 2.5 0 0 250 500 750 1000 1250 Time, Seconds A ~ V 4 YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CORE COOLANT LEVEL VS. TIME 411-10B TWO INCH DIAMETER BREAK

411:24 8/2f74 DRAFT O 8 I N L-N 4 m 8 8 8 ~ d b O g = 8 ed M owl l 'o O 4 A 5 5 5 5 uot43UId 'X2ttenD pinId k LOSS OF COOLANT ACCIDENT FIGURE YANKEE NUCLEAR POWER STATION HOT SPOT QUALITY VS. TIME 411-12A DECLG (CD=1.0)

411:25 8/2/74 DRAFT (V~') o i n 8 n e 8 a .!!s R O 8 4 g A f O 8 8 8 8 ~ ~ e e cos 7 J/mqi ' AlicoloA sseli 2 YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION HOT SPOT MASS VELOCITY VS. TIME 411-13A DECLG (CD=1.0) 1

411:26 8/2/74 DRAFT rsb o 8 i' i L m I o a m f G 8 E o L e m Ih y s esb .o \\ m To e5 / N 4 A ~_

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i 1 l C'\\ YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION HOT SPOT FILM COEFFICIENT VS. TIME 411-14A DECLG (CD=1.0)

411*.27 8/2/74 DF. AFT 8 8 E I 8 m C5 Y g o 2 / o 8 8 8 8 8 o n 8 C S cisd 'oanssaJd ) LOSS OF COOLANT ACCIDENT FIGURE YANKEE NUCLEAR POWER STATION CORE PRESSURE VS. TIME 411-15A DECLG (CD=1.0)

411'.28 8/2/74 DRAFT a E 8 4 8 8 m Ni: O a s e m e m n o 308/tH9I 0I ' MOI 3 Meo.zg p r\\ U YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION BREAK FLOWRATE VS. TIME 411-16A DECLG (CD= 1.0)

e a 411:29 8/2/74 DRAFT 8 ~ 8 e 8 8 m E i 8 2 t 2 o 8 W M, 8, pisa ' dojo oansso2d O YANKEE NUCLEAR LOSS OF C00LANI ACCIDENT FIGURE POWER STATION CORE PRESSURE DROP VS TIME 411-17A DECLG (CD=1.0)

l 411:30 8/2/74 DRAFT O 8 ~ 3 n ~ 8 A n 4 -s 8 8 0 r 2 m / 5 .S h m O L A g m 8 8 8 8 8 n E S 2c.radmaj. prI-) i l i YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION IIOT SPOT CLAD TEMPERATURE VS. TIME 411-18A DECLG (CD= 1.0) v

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411:34 8/2/74 ~ DRAFT ( ') i G' ,8 i . o t . ~, . o 4 8 8 1 ro ~ $y o x_/ E t l ~ o o o e o o o e o O S S O 8 8 Jos/mgr 'M Id S333 L (,) YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION ECCS FLOW (BLOWDOWN) 411-22A DECLG (CD=1.0) blhhhkkh

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411*.37 8/2/74 DRAFT O R i 4 m W 8 n ~ ? 8 e o* ^o S H m 6 O e i. m _= A N t N% AML " l o n 8 n 8 n A 5 uolacc.td 'X3TIcnD P!nid l O LOSS OF COOLANT ACCIDENT FIGURE YANKEE NUCLEAR POWER STATION HOT SPOT QUALITY VS. TIME 411-12B DECLG (CD=0.6) i t I

411".38 8/2/74 qV DRAFT 8 ~ R 8 n ~ m ? 8 8 e ~ 8 .g R O 8/ N / o 8 8 8 8 T "l ~ aas-y/mqt 'X2poloA sscw O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION HOT SPOT MASS VELOCITY VS TDfE 411-13B DFrtG (CD=0.61

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411'.43 8/2/74 DRAFT ,m N), 8 m ) 8 t rn E 8 8 er ~ .e e O \\ R h ? en t m A> O 8 8 8 e m m ~ d 'eangeaadural pelo i \\ O \\.) YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER ST*, TION 110T SPOT CLAD TEMPERATURE VS. TDiE 411-18B DECLG (CD =0.6)

411'.44 8/2/74 O DRAFT LJ 8 m W f ,5 s M i 8 l ~ 3e I .c e s-N J g '1 a 8 a 8 T m m d 'aangeaadtaal pinu O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION liOT SPOT FLUID TEMPERATURE VS. TIME 411-19B DECLG (CD=0.6)

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NUo R$ fU 0 0 0 2 5 7 1 0 5 1 5 2 1 s dnoc 0 e 0 S 1 em iT 0< 5 7 0 5 '8 3 5 2 / (v. t t h 0 5 0 5 0 5 2 o 2 0 7 5 1 1 0 0 0 $H n$ D sc 33 u. x O a8E szNm 99w e@*Sn8g4>nOE4 TCo Egx m 28 mS m$. eg4" <" s~E

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411'.53 I 8/2/74 DRAFT a f i i i o 2 8 z 5 i U O i m o' R ~ O } o c, I i l i' o o 8 8 8 8 8 m o m o w N N e4 M UTSd 'oanssaad I (v YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE 1 POWER STATION CORE PRESSURE VS. TIME 411-15C 0.6 x DECLS .l

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m 411:60 0 8/2/74 DRAFT -om - on i . a m n o 8 E m E - o t-- O - o M i o o o S S 8 8 o o o 5 e m e i I 30s/mqi ' mold S333 i ' O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION ECCS FLOW (BLOWDOWN) 411-22C 0.6 x DECLS

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411:67 8/2/74 " " ^ " O a ~ R 4 1 8 8 tn 0 R e ~ / o N C to e m N M C oas /inqr ot ' sold Meaag y O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE 411-16D POWER STATION BREAK FLOWpTE VS TIME 1.0 ft SI,0T

411:68 8/2/74 DRAFT AV o* a E 1 8 e J I H e O / o 4 d m o 8 4 n 8 e PTSd 'do20 oanssoad I YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CORE PRESSURE DROP VS. THIE 411-17D 1.0 ft2 SLOT

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XiAt, D Mq UUi YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CORE REFLOODING VS. TIME 411-21D 1.0 ft2 SLOT i

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411:75 8/2/74 bi WT l 50 - - l-T~ ~ -- ~ i 1 .q. 7 ... );!. _ t i I l._...._.. _. _ ..g y 1 l 40 - i I l i I i i 1 i l i i i i l i i i ee 30 i M 6 6 o' i l i . l. u i { l Pn i m 0 20 I i a. 't. o g -} e i 10 - I i u i i i l J._. j ._t .{ t l l i 4 1 0 0 40 80 120 160 200 Time, Seconds om o DW W U U. .ft O YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGURE POWER STATION CONTAINMENT PRESSURE VS. TIME 411-24D 1 n r,2gini . _ _.}}

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