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BNL-6 55 35

C S N F -7- 8 08/&-- SIMULATION OF SBWR STARTUP TRANSIENT AND STABILITY

ABSTRACT

Hsiang S . Cheng, Hasna IT. Khan, and Upendra S. Rohatgi

Brookhaven National Laboratory 4eCEIVED Upton, New York 11973

JUN 1 7 8 9 8 0 ST I

The Simplified Boiling Water Reactor (SBWR) designed by General Electric is a natural circulation reactor with enhanced safety features for potential accidents. It has a strong coupling between power and flow in the reactor core, hence the neutronic coupling with thermal-hydraulics is specially important. The potential geysering instability during the early part of a SBWR startup at low flow, low power and low pressure is of particular concern. The FL4MONA-4B computer code developed at Brookhaven National Laboratory (BNL) for the SBWR has been used to simulate a SBWR startup transient and evaluate its stability, using a simplified four-channel representation of the reactor core for the thermal-hydraulics. This transient was run for 20,000 sec (5.56 hrs) in order to cover the essential aspect of the SBWR startup. The simulation showed that the SBWR startup was a very challenging event to analyze as it required accurate modeling of the thermal-hydraulics at low pressures. This analysis did not show any geysering instability during the startup, following the startup procedure as proposed by GE.

1. INTRODUCTION

The Simplified Boiling Water Reactor (SBWR) designed by General Electric is an advanced passive design using natural circulation for coolant flow without any active pumps. Details of the design for various safety features are available in the Standard Safety Analysis Report ( S S A R ) of the SBWR [ 11. The startup procedure of the SBWR is of special importance since the low pressure, low flow and low power prevailing in the early part of the transient can become a precursor to an instability. In a natural circulation system like the SBWR, the core flow is strongly coupled to the reactor core power. Startup of the reactor begins with slow heat-up of the coolant in the reactor followed by removal of selected groups of control rods. Gradual increase of reactor power results in single-phase natural circulation flow within the reactor, while the system pressure is still very low. During this period, transition fi-om a subcooled core to a saturated core with a subcooled chimney may be accompanied by creation of large vapor bubbles which can initiate geysering type instability. Certain combination of pressure, flow and thermal-hydraulic conditions was required for geysering to occur in small scale laboratory experiments such as Aritomi et al. [2,3] and Wang et al. [4]. According to these research findings and supporting analyses by Paniagua et. Al. [5 ] , out of phase geysering in two parallel channels is limited to low-pressure and low-flow conditions. A subcooled chimney allows condensation of the large bubbles leading to instability in the flow within the channels. At higher power, a loop type instability may follow the geysering instability, where the

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i

flow in the downcomer and the core oscillates in phase. Finally, density wave oscillations of higher frequencies can occur at higher power-to-flow conditions.

Although oscillations during startup has been demonstrated in the laboratory scale experiments, the two dt~&&$a$%

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or use- fulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any spe- cific commercial product, process, or service by trade name, trademark, manufac- turer, or otherwise does not necessarily constitute or imply its endorsement, Tccom- mendation. or favoring by the United States Government or any agency thereof. The views and opinions of authors expnssed henin do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER

Portions of this document may be illegible electronic image products. Images are produced from the best available original document.

c . Table 1. Summary of SBWR Startup Transient Phases

Phase I Phase I1 1.1 I I11 IV

Deaeration I 5 6 7 8 4 Heating 0 n S e t of S a t r a t i n U 0 Pressurization Auto FW Control Power rise to 100% 10% Power M i n Turbine Synchronization Npch CRD Flow Level in

C h i m n e Y

i n 1 e t

Auto FW Control

Steam through bypass valve

Pressure Regulator

Large Low Ahi Void Low Power

Density

Oscillation

Power%< 2% Flow % < 12%

TI = Onset of NVG T2 = Saturation in Chimney Inlet T3 = Min Npch reached or Possibility of Density Wave

2.1.2 Saturation at the Inlet to the Chimney

the saturation temperature in the chimney varies along its height of 10 m, with the lowest at the top. During Phases I and 11, the coolant in the chimney remains subcooled. Migration of the vapor bubbles to the chimney results in condensation in the subcooled liquid and a consequent heating of the coolant. This process of heat transfer to the subcooled fluid in the chimney through bubble condensation continues and the fluid in the chimney gradually reaches saturation at different axial levels. The transition from Phase I1 to Phase 111 takes place when the fluid at the bottom of the chimney is saturated.

2.1.3 Minimum Npch Condition

As the startup proceeds, the control rods are withdrawn and core power increases. The chimney is in saturation state with vapor generation taking place and the reactor may go into loop type oscillations. Lahey [9] has described a stability map for a BWR based on linear stability analysis as shown in Figure 1. The abscissa represents the Phase Change Number (Zuber Number) Npch, which is proportional to power. The ordinate represents the Subcooling Number, Nsub whch is proportional to core inlet subcooling. As Figure 1 indicates, there is a minimum Npch below which the reactor is stable with respect to density waves. Thus, the transition between phases 111 and N can be defined in terms of the minimum Npch.

2.2 Simulation of the SBWR Startup Using the RAMONA-4B Code

2.2.1 The RAMONA-4B Code

The SBWR startup transient has been simulated using the RAMONA-4B code developed at BNL [10,11]. The code models all the important components in the reactor pressure vessel (RPV), such as the reactor core, downcomer, lower plenum, upper plenum and riser, jet or internal pumps, steam separators and dryers, and the steam dome, as well as the control and plant protection systems, steam lines, balance of plant (BOP), and containment. SBWR specific components such as the isolation condenser (IC) and the standby liquid control system (SLCS) have also been modeled. The neutron kinetics is modeled with a time-dependent 3D diffusion theory with one and half group of prompt neutrons and six groups of delayed neutrons for a maximum of 804 neutronic channels each with a maximum of 24 axial nodes. Local thermal-hydraulic feedback is taken into account in terms of the changes in nodal two-group cross sections due to the local void, fuel temperature, and moderator temperature. The two-phase thermal-hydraulics in the core is modeled via a drift-flux formulation with flow reversal capability for nonequilibrium, nonhomogeneous flow through multiple (up to 200) parallel coolant channels.

The SBWR specific models implemented into RAMONA-4B are the isolation condenser (IC), standby liquid control system (SLCS), and local boron transport. The boron transport model solves the boron transport equation in each of thermal-hydraulic cells in the RPV by means of standard donor-cell differencing with flow reversal logics. The detailed description of the various RAMONA-

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4B models is given in Ref. [lo] and that of its input descriptions is provided in Ref. [l 13. Here we point out the following important models required for a credible simulation of the SBWR startup transient:

(1) (2) (3) (4)

(5 ) (6)

Accurate low-pressure properties fiom 0.03 bar to 40 bar. A heat slab model for stored energy in structure materials. A flow-dependent loss coefficient model. A Reactor Water Cleanup and Shutdown Cooling System (RWCWSDC) model for level control during the SBWR startup. A control rod drive (CRD) flow model. A turbine bypass control model at pressures higher than 19 bar.

These models have been implemented in the RAMONA-4B code.

It should be mentioned that the heat transfer models for boiling and condensation at low pressures are not well established at present in the literature. Furthermore, the accuracy of these models at low pressures has not been validated for RAMONA-4B. Since the accuracy of these models may have a direct impact on the predicted phase change, thermal and flow characteristics, the results presented here should be considered qualitatively only.

2.2.2 Calculation Model for the Startup Transient

A simulation of startup conditions for the SBWR has been performed at BNL using the RAMONA- 4B code. Since information on the spatial and temporal movements of the control rods are not available fiom the GE report, a time-dependent power ramp profile as proposed by GE has been imposed as a boundary condition and a thermal-hydraulic only calculation performed with a simplified core model. The simplified core model is necessitated by the need to run the startup transient over a long period of time (5-6 hours) in order to show the important characteristics of the early part of the startup transient. The simplified core model consists of four parallel coolant channels including the bypass channel. The chimney, steam dome, downcomer, and the lower plenum were also included. Table 11 presents the geometric parameters and nodalization scheme of the RAMONA-4B input deck.

Table I1 RAMONA-4B Input Deck for the Startup Transient

Component D.comer 1 D.comer2 Lower P1.1 Lower P1.2 Core Riser Dome

No. of Nodes 9 3 2 3 18 6 1

Height (m) 10.58 4.78 2.14 2.05 2.74 10.58 6.95

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2.2.3 Initial Conditions for Startup

The initial conditions were selected to be as close to the actual values proposed by GE. Table I11 shows the initial conditions used in the present simulation.

Table I11 Initial Conditions for the Startup Transient

Pressure Coolant Temp. Core Power Feedwater Coolant Level (MPa) ("C) (Mw) Flow (m)

(KgM 0.055 40 0.02 0.0083 0.655*

* Above the entrance to upper downcomer (1 8.17 m above core bottom)

2.2.4 Boundary Conditions for Startup

Estimation of certain boundary conditions were made following the results of T U C G by GE. Table IV presents the boundary conditions used for the startup simulation.

Table IV Boundary Conditions for the Startup Simulation

I Boundary Condition I Input to RAMONA-4B I Core Power

Steam Flow

Feed water Temperature

A ramp power of 20 M W h was imposed as a boundary condition, which corresponds to 42 "Ch.

Turbine Control Valve (TCV) remained closed during the transient allowing reactor pressure to rise.

Feed water heaters were inoperative since TCV was closed during the transient. Feed water temperature remained fixed at 80°C.

The most important boundary condition used for the present simulation was a time-dependent power ramp profile as proposed by GE shown in Figure 2.

2.3 Simulation Results

The startup transient was run for 20,000 s (5.56 h) in order to show the important characteristics of the early part of the startup. Here we present the most important results to illustrate the basic characteristics of the SBWR startup transient.

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2.3.1 Pressure Response During Startup

Pressure rise in the steam dome during startup is a result of heating by the power ramp, level swell during expansion, and net vapor generation in the reactor. Figure 3 shows the result obtained for the pressure response during the startup. Due to low expansion rate for the liquid of the liquid, the pressure rise is negligible during Phase I. Onset of net vapor generation in the core is the transition between Phase I and Phase II. The present results indicate that sustained bubble concentration begins approximately at this transition time. Since the turbine control valve (TCV) and the bypass line remained closed, this resulted in additional rise in pressure.

2.3.2 Core Flow Response During Startup

Single phase natural circulation in the core results in core flow at a low power and low pressure. For a given power, the predicted flow is dependent upon the core inlet loss coefficients for the parallel coolant channels. The total core inlet flow during the startup transient is shown in Figure 4. The total core flow comprises of the sum of the flows through the 3 heated channels and the unheated bypass channel as shown in Figure 5. The flow rates through the two larger channels are equal, while the flow through the orificed channel is less than that of the bypass channel. Change in liquid density during heat-up causes the lighter fluid to rise upward. Recirculation of the hot fluid from chimney through the downcomer results in a continuous rise in coolant temperature. The flow rates through the coolant channels increase rapidly at the onset of vapor generation in the core due to the enhanced buoyancy forces. This is caused by the large density change accompanied with phase change occurring at 5400 s. However, the flow starts to decrease around 6000 s as the void begins to collapse in the core following the pressure rise. The qualitative nature of the results is acceptable; however, the quantitative aspect would require further assessments of the vapor generation and condensation models for low pressure conditions.

2.3.3 Coolant Temperature Response During Startup

Increase in the coolant temperature during startup is monotonic with time. There is an axial gradient in temperature as shown in Figure 6. The coolant temperature in the core follows the saturation temperature at that axial location after the onset of net vapor generation at approximately 5400 s. Since the system pressure increases with time, the coolant temperature also follows this behavior.

2.3.4 Void Response During Startup

Figure 7 shows the void hction prediction for the startup transient. Subcooled vapor generation in the core begins in Phase I. Since the bulk fluid in the channels remain subcooled in this phase, condensation of the vapor takes place until there is net vapor generation. The void fi-action in the core increases rapidly with time until a rise in system pressure accompanied by increase in saturation temperature causes some of the vapor to condense. The magnitude of core averaged void fraction shown in Figure 7 is very small. Thus, a small change in saturation temperature can cause condensation of the vapor. In addition, the increase in the core flow rate causes the vapor to be

carried away from the core region. This is evident in Figure 7 beyond 6000 s of the transient. Depending upon the rate of vapor generation due to the wall heat transfer and the degree of subcooling in that vicinity, the net vapor generation will be noticeable. Figure 8 shows the net vapor generation rate in the core. It begins at approximately 5400 s and continues to rise until about 14,000 s. The pressure rise caused by this vapor generation, the level swell and the increase in flow results in a rapid decrease in the void fraction.

2.3.5 Collapsed Liquid Level Response During Startup

It is important to maintain a desired liquid level in the steam dome of the vessel during the startup. This is done through the RWCU and SDC systems. Figure 9 shows the collapsed liquid level response predicted by RAMONA-4B. It is seen that the RWCU and SDC models in RAMONA-4b was able to maintain a constant liquid level in the vessel during the startup except for some mild oscillations between 5400 s and 9000 s at time of rapid increase in vapor generation in the core.

2.3.6 Flow Oscillation

Figures 7,s and 9 also indicate some form of oscillation between 5600 s and 8000 s. These oscillations have a period of 800 s and they disappear as the system pressure increases. These oscillations are a result of loop type instability. Increase in void fraction leads to larger buoyancy force and core flow. The larger core flow will decrease the void fraction and buoyancy force leading to reduced core flow. This process continues until the system pressure is large enough to supresss any large increase in the buoyancy forces due to density difference between the phases.

3. A STUDY OF SBWR STARTUP INSTABILITY

Since neither the sustained geysering instability nor the sustained loop type instability were observed in the SBWR startup simulation by RAMONA-4B, an attempt was made to find conditions which will lead to sustained oscillations by imposing an abnormal (but unrealistic) power ramp profile as the boundary condition. A rapid increase of power with a ramp fkom 20 kW to 140 MW within 0.2 s was specified through the input. The startup transient was run as in the base case presented in Section 2, except that it was run for only 3000 s when a sustained oscillation has been established. The pressure was held constant once it reached 0.19 MPa at which time oscillations in the core flow began. This combination of pressure, flow and power, where TRACG predicted oscillations [ 1 11, was identified as an unstable region by GE. It is claimed by GE that, during a normal startup of the SBWR, the reactor is not expected to attain this condition.

The predicted results by RAMONA-4B are presented in Figures 10 through 13. Figures 10 and 1 1 show the pressure and core inlet flow responses, respectively. Since the magnitude of the oscillation is small, an amplification of the flow oscillation is made to reveal its amplitude and frequency as shown in Figure 12, where the oscillation was found to continue in time with a fixed amplitude of 47 kg/s and a period of 42.4 s. According to Figure 13, the net vapor generation rate oscillates with the same fkequency as a result of consecutive vaporization and condensation. The oscillations are

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therefore condensation induced. Because of the uncertainty of the condensation and vapor generation models in RAMONA-4B for low pressures, these results should be considered as qualitative.

Core Inlet Loss Coefficient

67.9

44.5

21.4

12.2

Due to the well known effect of the loss coefficient on the BWR stability, the effect of core inlet loss coefficient on the core flow oscillation was investigated. It should be noted that the effect of the core exit and chimney loss coefficients on the flow oscillation is negligible because of the extremely low vapor fiaction in the core and chimney (see Figure 8). Therefore, the investigation did not consider the variation of the core exit and chimney loss coefficients. Several cases were run with the same boundary conditions and different initial flow conditions. The steady state flow for a given power corresponds to the core inlet loss coefficient as shown in Table V. The amplitude of flow oscillation is found to increase with decrease in the core inlet loss coefficient, which corresponds to the stability of the single phase flow at the core inlet. The initiation of the flow oscillation occws between 1520 s to 1600 s, which correspond to the pressure reaching 0.19 MPa in all the cases. The period of oscillations ranges fiom 31.8 s to 46.7 s. The GE simulation using TRACG has predicted flow oscillations of a period of approximately 50 s with a larger amplitude. The peak void fiaction of RAMONA4B calculation is less than that of TRACG, which is due to the differences in the condensation and vapor generation models.

Initiation Time of Amplitude of Period of Oscillation Average Cor Flow Oscillation (s) Oscillation (kgh) 6) Rate (kg/s)

1600 38 46.7 2600

1580 47 42.4 3000

1540 64 35.9 3700

1520 76 31.8 4200

Table V Effect of Core Inlet Loss Coefficients on the Abnormal Startup

4. SUMMARY AND CONCLUSIONS

Startup of the SBWR begins at a low pressure of 0.055 MPa and ends at full pressure of 7.2 MPa. Successful application of RAMONA-4B in this wide pressure range has been demonstrated. The most important modeling requirements for the startup transient are the low pressure properties, the heat slab model for the stored energy, and the level control via the RWCU/SDC systems. Accurate modeling is also required for condensation and vapor generation at low pressures. Unfortunately, data for low pressures are very scarce at best in the open literature. The accuracy of the existing models in RAMONA-4B is unknown at present and needs assessments against experimental data, if any. However, we believe that the trend of the predictions by RAMONA-4B are qualitatively correct for the SBWR startup transient.

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Neither the sustained geysering instability nor the loop type instability were predicted by RAMONA-4B in the startup simulation following the recommended procedure by GE. The density wave oscillation was not observed at all because the power level used in the simulation was not high enough. A study was made of the potential instability by imposing an unrealistically high power ramp in a short time period as suggested by GE. Core flow oscillations of small amplitude were predicted by RAMONA-4B with a period of between 3 1.8 s and 46.7 s similar to that of the TRACG prediction by GE.

ACKNOWLEDEMENTS

This work was performed under the auspices of the U.S. Department of Energy, Contract #DE- AC02-98CH10886. The author is a retired Brookhaven National Laboratory employee.

5. REFERENCES

1. General Electric Nuclear Energy, “SBWR Standard Safety Analysis Report,” GE Report No. 25A5113, Rev. A, 1992.

2.

3.

4.

5 .

6.

7.

8.

M. Aritomi, T. Nakahashi, J. H. Chiang, M. Wataru, and M. Mori, “Transient Behavior of Natural Circulation for Boiling Two Phase Flow (Experimental Results)”, 6th Proc., Thermal hydraulics, A N S Winter Annual Meeting, Washington D.C., 1990.

M. Aritomi, J. H. Chiang, and M. Mori, “Geysering in Parallel Boiling Channels”, Nuclear Engineering and Design, Vol. 14 1, pp 1 1 1 - 12 1, North Holand Physics Publishing, 1 993.

S. B. Wang, J. Y. Wu, C. Pan, and W. K. Lin, “Thermal Hydraulic Oscillations in a Low Pressure Two-Phase Natural Circulation Loop at Low Powers and High Inlet Subcooling,” 4th International Topical Meeting on Nuclear Thermal Hydraulics, Operations and Safety, Taipei, Taiwan, 1994.

J. Paniagua et. al;” Thermal Hydraulic Instabilties During Start-up in Two Heated Parallel Channels”, ASME Winter Annual Meeting, Dallas, Texas, 1997

W. H. M. Nissen et al., “The Startup of the Dodewaard Natural Circulation BWR Experience,” International Conference on Design and Safety of Advanced Nuclear Power Plants, Vol. 3,1992.

“TRACG Qualification Report,” NEDE 321 77P, General Electric, 1993.

U.S. Rohatgi, et. al.; “PIRT for SBWR Plant Start-up Stability”, NUREG/CR 6474, USNRC and Brookhaven National Laboratory Technical Report, June, 1995

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9.

10.

11.

R. T. Lahey, Jr. and F. Moody, “The Thermal-hydraulics of a Boiling Water Nuclear Reactor,” Second Edition, A N S Publication, 1993.

U. S. Rohatgi, H. S. Cheng, H. J. Khan, and L. Y . Neymotin, “RAMONA-4B: A Computer Code with Three-Dimensional Neutron Kinetics for BWR and SBWR System Transients,” Vol. 1 : Models and Correlations, NUREGICR-6359, BNL-NUREG-52471, Brookhaven National Laboratory, 1997.

U. S. Rohatgi, H. S. Cheng, H. J. Khan, and L. Y . Neymotin, “RAMONA-4B: A Computer Code with Three-Dimensional Neutron Kinetics for BWR and SBWR System Transients,” V01.2: User’s Manual, NUREGKR-6359, BNL-NUREG-5247 1, Brookhaven National Laboratory, 1997.

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Figure 1. Typical B W 4 Stability Map & = 27.8, Kexit = 0.14) (Ref. Lahey & Moody, 1993, Fig. 7-10).

SBWR STRRTUP TRRNSIENT WITH 4-MNNEL FULL CORE: MOOEL or EOC CORE THERNRL POWER

RWCU Level ControL (FRWC-1.21, No Pressure Control by Byposs

7.0

0 . 0 y . 0.0 i . 0 4.0 6.0 8.0 10.0 12.0 14.0 18.0 18.0 1.0 ...

TIME [SI xlOJ

Figure 2. Time-Dependent Power Profile as a Boundary Condition.

12

SBWR STARTUP TRANSIENT WITH +CHANNEL FULL CORE MODEL o t EOC RWCU LeveL Control. (FRWC-1.21, No Pressure ControL by Byposs

RVERRGE VESSEL PRESSURE n

E! I 50.0

- U 0 W [r 3 m m W LT e

.O TIME (SI *lo’

Figure 3. System Pressure Response During the Startup Transient.

SBWR STARTUP TRANSIENT WITH 4-CHANNEL FULL CORE MOOEL o t EOC RWCU Levet ControL IFRWC-1.21, No Pressure ControL by Byposs

TOTRL CORE INLET MASS FLOW RRTE

30W.O

xw.0 -

2MNl.O- - UI \ m 1 IMo.0-

LJ c

b I

a 1wo.o-

x a -I L

sw.0 - 0.0-

-5W.01 . . I . . I 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0

TIME (SI

Figure 4. Total Core Inlet Flow Response During the Startup Transient.

13

Iom.0

9m.o

8m.o

7m.o

6W.O

5m.0

4m.o

3m.o

2 m . O

1m.o

0.0

-1m.o

SBWR STARTUP TRRNSIENT WITH 4-CHANNEL FULL CORE MOOEL at EOC RWCU LeveL ControL (FRWC-1.21, No Pressure ControL by Byposs

INLET MRSS FLOH RRTES AT CHANNEL 1, 2, 3, ond 4

I \.-

0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 TIME ( 5 )

.O ,a’

Figure 5. Channel Inlet Flow Responses During the Startup Transient.

SBWR STARTUP TRRNSIENT WITH 4-CHRNNEL FULL CORE MODEL at EOC RWCU LeveL ControL (FRWC-1.21, No Pressure ConLroL by Byposs

LIOUIO TEMPERATURES IN REACTOR CORE 275.0 >

- 0 w LL 3 I-

Lz W a

I-

a

6

25.0 . . , .. TIME IS1 x1O’

Figure 6 . Coolant Temperature Response During the Startup Transient.

14

I .

1

P

SBWR STARTUP TRRNSIENT WITH 4-CHRNNEL FULL CORE IIODEL ot EOC RWCU LeveL ControL IFRWC-1.21, No Pressure ControL by Byposs

CORE RVERRGE VOIO FRACTION

K

7 5.0 8.0 10.0 12.0 14.0 16.0 18.0 I

TlHE IS1 *loJ

Figure 7. Core Average Void Response During the Startup Transient.

SBHR STRRTUP TRANSIENT WITH 4-CHANNEL fULL CORE MODEL ot EOC RWCU LeveL ControL (FRWC-1.21, No Pressure Control. by Byposs

TOTRL VRPOR GENERRTION RATE IN RPV

v) \ CI) Y

W c U LL

z 0 c- LL

W LF)

LL 0 Q U >

-

Y

-0.01 ! . , . . . , . . .’. , . . , . . . I . . . . , . . . . , 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 0

TIME (SI .lo’

Figure 8. Total Vapor Generation Rate Response During the Startup Transient.

15

5

4.04 , I , , , , , , , * , I 0.0 2sn.o Xa.0 m.0 Iae.0 12sn.o I ~ . O m4.0 m . 0 pI0.0 zsm.0 m.0 am.0

TIHE 1st

Figure 10. Reactor Pressure vs. Time During Abnormal Startup.

16

* I

SBWR STARTUP TRANSIENT WITH 4-CHANNEL FULL CORE MOOEL at EOC RWCU LeveL ControL [FRWC-1.21, No Pressure Control by Bypass

COLLRPSED WATER LEVEL RESPONSE 3.0

2.0 -

1.0-

0.0- .

-1.0-

-3.0 f 0.0 2.0 4.0 6.0 8.0 10.0 12.0 1i .o 16.0 18.0 i 1.0

TIME I S 1 w10'

Figure 9. Collapsed Liquid Level Response During the Startup Transient.

No RwRl LwmL ConlroL, Prommurr Hold C m a L a L 01 1.9 b- RVERRGE VESSEL PRESSURE

b B.0 I

1 5 m . O

2250.0

aw.0

150.0

0.0 0.0 250.0 y10.0 M.0 lom.0 llso.0 Irm.0 I M . 0 mD.0 22so.o am.0 mo.0 :

TIM (SI 'LO

Figure 1 1. Core Inlet Flow Rate vs. Time During Abnormal Startup.

No RIXXl Lsvd CmlroL Prossun Held ConslonL 01 1.9 B a r MRE MluvkL INLET WSS I7.W RATE

1.0

Figure 12. Magnification of Core Inlet Flow Rate Oscillations During Abnormal Startup Simulation.

17

Na Rull LwoL Cmtrd Proosuro Hold ConsLmL o t 1.9 Ba TOT% 'id GMERRTION RATE I N RPV

0.m 0.0 an.0 5m.o m.0 1m.o 1an.o Irm.0 I*O ooo.0 p50.0 a . 0 m.0

T I E IS1 0.0

Figure 13. Net Vapor Generation Rate vs. Time During Abnormal Startup Simulation.