NUREG/IA-0180

International 

Agreement Report 

Application of RELAP5/MOD3.1 

to ATWS Analysis of 

Control Rod Withdrawal 

From 1% Power Level 

Prepared by 

H. Kantee 

Fortum Engineering Ltd. 

RO. Box 10 

00048 Fortum 

FINLAND 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

June 2000 

Prepared as part of 

The Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0180

~ International 

Agreement Report 

Application of RELAP5/MOD3.1 

to ATWS Analysis of 

Control Rod Withdrawal 

From 1% Power Level 

Prepared by 

H. Kantee 

Fortum Engineering Ltd. 

P.O. Box 10 

00048 Fortum 

FINLAND 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

June 2000 

Prepared as part of 

The Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
ABSTRACT 

The RELAP5/MOD3.1 computer code has been applied in the analysis of a regulating control rod group withdrawal from 1 % power level. The analysis has been carried out 

as an Anticipated Transient Without Scram (ATWS) event. 

The analysis is related to an extensive inherent boron dilution study which was performed to Loviisa Nuclear Power Plant (NPP) in the mid 1990's. The main goal of the analysis was to study if during the accident so called boiling-condensing mode develops and thereafter a water plug with low boron concentration is formed in the 

cold leg loop seal.

i
TABLE OF CONTENTS 

ABSTRACT 

TABLE OF CONTENTS 

LIST OF TABLES 

LIST OF FIGURES 

EXECUTIVE SUMMARY 

1 INTRODUCTION ................................................................................................................................................... 1 

2 NODALIZATION M ODEL .................................................................................................................................. 2 

3 ASSUM PTIONS AND ANALYZED CASES .................................................................................................. 3 

4 ANALYSIS RESULTS ........................................................................................................................................... 5 

4.1 BASE CASE, REACTIVITY OF WITHDRAWAL 3.6 $ ............................................................................................... 5 

4.2 BASE CASE, REACTIVITY OF WITHDRAWAL 4.0 $ ............................................................................................ 6 

4.3 REACTIVITY OF WITHDRAWAL 2.2 $ ..................................................................................................................... 6 

4.4 REACTIVITY OF WITHDRAWAL 1.4 $ ..................................................................................................................... 7 

4.5 REACTIVITY OF WITHDRAWAL 1.8 $ ........................................................... ........... .................. 7 

4.6 REACTIVITY OF WITHDRAWAL 1.9 $ ..................................................................................................................... 8 

4.6.1 Reactivity of withdrawal 1.9 $ and turbine bypass not available .......................................................... 9 

4.6.2 Reactivity of withdrawal 1.9 $ and one main feedwater pump available ............................................. 9 

5 RUN STATISTICS .............................................................................................................................................. 10 

6 CONCLUSIONS ................................................................................................................................................. 10 

REFERENCE 

TABLES 

FIGURES 

APPENDIX

ii

I I
LIST OF TABLES

Table 2-1 Initial value of the main parameters 

Table 2-2 Initial control rod position and reactivity worth 

Table 5-1 Run time statistics

iii
LIST OF FIGURES

Figure 2-1 

Figure 2-2 

Figure 2-3 

Figure 2-4 

Figure 2-5 

Figure 4.1 -1 

Figure 4.1-2 

Figure 4.1-3 

Figure 4.1-4 

Figure 4.1-5 

Figure 4.1-6 

Figure 4.1-7 

Figure 4.1-8 

Figure 4.1-9 

Figure 4.1-10 

Figure 4.1-11 

Figure 4.1-12 

Figure 4.2-1 

Figure 4.2-2 

Figure 4.2-3 

Figure 4.2-4 

Figure 4.2-5 

Figure 4.2-6 

Figure 4.2-7 

Figure 4.2-8 

Figure 4.2-9 

Figure 4.2-10 

Figure 4.2-11 

Figure 4.2-12 

Figure 4.3-1 

Figure 4.3-2 

Figure 4.3-3 

Figure 4.3-4 

Figure 4.3-5 

Figure 4.3-6 

Figure 4.3-7

Pressure vessel nodalization 

Loop YA11 nodalization 

Pressurizer nodalization 

HPI, make-up, letdown and purification systems nodalization 

Steam generator YB 11, steam line and feedwater injection system 

nodalization 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part 

Total reactivity 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part 

Total reactivity 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

iv

I I
Figure 4.3-8 

Figure 4.3-9* 

Figure 4.3-10 

Figure 4.3-11 

Figure 4.3-12 

Figure 4.4-1 

Figure 4.4-2 

Figure 4.4-3 

Figure 4.4-4 

Figure 4.4-5 

Figure 4.4-6 

Figure 4.4-7 

Figure 4.4-8 

Figure 4.4-9 

Figure 4.4-10 

Figure 4.4-11 

Figure 4.4-12 

Figure 4.5-1 

Figure 4.5-2 

Figure 4.5-3 

Figure 4.5-4 

Figure 4.5-5 

Figure 4.5-6 

Figure 4.5-7 

Figure 4.5-8 

Figure 4.5-9 

Figure 4.5-10 

Figure 4.5-11 

Figure 4.5-12 

Figure 4.6-1 

Figure 4.6-2 

Figure 4.6-3 

Figure 4.6-4 

Figure 4.6-5 

Figure 4.6-6 

Figure 4.6-7 

Figure 4.6-8 

Figure 4.6-9 

Figure 4.6-10 

Figure 4.6-11

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part Total reactivity 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part 

Total reactivity 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part 

Total reactivity 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part

v
Figure 4.6-12 

Figure 4.6.1-1 

Figure 4.6.1-2 

Figure 4.6.1-3 

Figure 4.6.1-4 

Figure 4.6.1-5 

Figure 4.6.1-6 

Figure 4.6.1-7 

Figure 4.6.1-8 

Figure 4.6.1-9 

Figure 4.6.1-10 

Figure 4.6.1-11 

Figure 4.6.1-12 

Figure 4.6.2-1 

Figure 4.6.2-2 

Figure 4.6.2-3 

Figure 4.6.2-4 

Figure 4.6.2-5 

Figure 4.6.2-6 

Figure 4.6.2-7 

Figure 4.6.2-8 

Figure 4.6.2-9 

Figure 4.6.2-10 

Figure 4.6.2-11 

Figure 4.6.2-12

Total reactivity

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part 

Total reactivity 

Total and fission power 

Core entrance, core exit and saturation coolant temperature 

Collapsed liquid level above hot leg center line, in core and in 

pressurizer 

Void fraction in core volumes 

Void fraction in hot leg 

Core entrance and bypass mass flow rate 

Pressure in hot leg, cold leg and pressurizer 

Primary coolant mass inventory 

Steam generator liquid level 

Boron concentration core volumes 

Boron concentration in cold leg loop seal horizontal part 

Total reactivity

vi
EXECUTIVE SUMMARY 

An extensive study for Loviisa NPP concerning inherent boron dilution was carried out in the mid 1990's. One of the main concerns was that during a transient or an accident liquid level in the pressure vessel decreases below the hot leg elevation and so called boiling-condensing mode begins. Boiling-condensing mode means that coolant boils in the pressure vessel and steam condenses in the steam generators forming a low boron concentration water plug in the cold leg loop seal. In case of Loviisa, condensing in the steam generator is obvious, since primary side cooldown is done by cooling down the secondary side. 

One of the accidents which was analyzed during the study using RELAP5/MOD3.1 computer code was a regulating control rod group withdrawal from 1 % power level 

RELAP5/MOD3.1 

as an ATWS case. Concerning this kind of analysis the point core kinetics model in is not a very good model to calculate power behavior since it does not take into consideration the change in the axial power profile when the control rod is moving upward. However, from the boron dilution point of view this shortcoming was not considered significant because the main goal was to find out events during which boiling-condensing mode develops. Therefore RELAP5/MOD3.1 code was used in the preliminary analysis to get qualitative data, which was later utilized when the sequences were considered for detailed analyses using the coupled thermal hydraulic and 3-D reactor dynamics code HEXTRAN [1]. 

The RELAP5/MOD3.1 analyses results showed that continuous long-term boiling condensing mode did not develop. Instead it was typical that one or sometimes more boiling-condensing mode phases took place during an accident but they lasted so short time that a serious boron dilution problem was not discovered. The maximum boron dilution was only 300 ppm, which corresponds to 2.3 % reactivity. This kind of dilution would probably not cause a serious core power transient during natural circulation even if all the control rods would be above the core.

vii
I INTRODUCTION

The first scenarios of local boron dilution accidents concerning Loviisa NPP were 

analyzed in 1990. It was quite soon recognized that the loop thermal-hydraulics in the 

two-phase mode had a strong effect on the development of inherent boron dilution 

conditions. Therefore, scoping studies of different ATWS sequences were also made 

using RELAP5 code with the point kinetics core model. 

The inherent boron dilution arises in the primary circuit when boiling-condensing 

cooling mode prevails. Boiling-condensing mode occurs when coolant boils in the 

pressure vessel and steam condenses in the steam generators forming a low boron 

concentration water plug in the cold leg loop seal. Very detailed description of the 

thermal hydraulics and core neutronics is necessary to reliably analyze the event. It 

was found that only 3-D modeling of neutronics is capable to predict the core 

reactivity level reliably in all conditions because axial and even radial density 

distributions in the core strongly changed during the transient due to different flow 

conditions: nominal forced flow, main circulation pump rundown, and different 

natural circulation phases. Using fewer dimensions in the core modeling in some cases 

changed the whole accident scenario with respect to the occurrence of inherent boron 

dilution. 

The amount of the primary coolant mass lost during the pressure increase and after 

stop of the reactor coolant pumps (RCP) depends crucially on the fission power 

behavior of the core. In the beginning of cycle (BOC) conditions, when the reactivity 

feedback of coolant density is very weak, there must be boiling in order to decrease 

the fission power, and the coolant mass inventory can decrease to such low level that 

boiling-condensing mode occurs. In the middle of cycle (MOC) conditions, threat of 

inherent boron dilution during ATWS cases is over because of clearly stronger 

feedback of coolant density. The studies with the different core models indicated that 

reliable fission power prediction during the transient was only achieved with a 3-D 

core model describing neutronics and flow channels of the fuel assemblies 

individually. Using less dimensions in the core modeling did in some cases change the 

whole accident scenario with respect to the occurrence of inherent boron dilution, 

even though the model was separately validated both in nominal and natural 

circulation flow conditions. 

Mixing provides an effective natural defense against severe reactivity excursions. 

Mixing occurs already in the steam generator in the build up phase of the diluted 

water plug and later in the reactor pressure vessel downcomer and in the lower 

plenum, when the diluted plug starts to move towards the core. Consequences of a 

reasonably conservative 50 % diluted water plug reaching the core with natural 

circulation would be limited to heat transfer crisis and consequent temporary fuel 

cladding overheating. High fuel temperatures are not to be expected with natural 

circulation. 

It should also be emphasized that a common cause failure causing mechanical 

jamming of all control rods in the VVER-440 reactor is considered to be highly

I
improbable. Drop of at least some of the rods help the shutdown of the core to such an 

extent that natural circulation is not disturbed. In the case of electrical failure even 

delayed scram actuated by the operator effectively eliminates the consequences of 

possible diluted water plugs. 

In 1996 RELAP5/MOD3.1 computer code was applied in the ATWS analysis of a 

regulating control rod group withdrawal from 1 % power level. When RCPs are not 

running and the coolant inventory is reduced in the primary circuit, it is possible that 

two-phase natural circulation is interrupted when liquid level decreases below hot leg elevation. The main goal of the analysis was to investigate if during the accident this 

phenomenon takes place and a boiling-condensing mode develops. If it does, there is a 

possibility that steam generated in the core flows to the steam generators, condenses 

and forms a diluted water plug in the cold leg loop seal. The analysis concentrated on 

the formation of the diluted plug and less attention was paid on the reactivity effect 

when the diluted plug flows into the core after restart of the two-phase natural 

circulation. 

The shortcoming of the core model in the RELAP5/MOD3.1 code in this kind of 

event is that it cannot calculate the change in axial power profile when a control rod 

moves upwards and when a diluted water plug flows into the core. On the other hand, 

the effect of the control rod movement can be taken into account in the point kinetics 

model in such a way that the total power generation in the core is tuned to correspond 

the real power generation. Control rod group withdrawal is given in the code input so 

that the insertion of reactivity is based on the elevation where the withdrawal starts 

and it changes linearly as a function of control rod position. Additionally, also 

reactivity feedback from coolant density, fuel temperature and boron concentration is 

needed. 

Several parameter variations were carried out and the main interest was focused on the 

reactivity worth of the control rod group withdrawal, which mainly depends on the 

initial position of the control rod. Therefore, because of strong reactivity variations 

from the control rod group withdrawal, the feedback inaccuracy of point kinetics 

model from coolant density and fuel temperature at least partly disappeared. 

The selection of RELAP5/MOD3.1 code to these analyses can be justified since the 

main goal of the analyses was to get qualitative data, which was later utilized when 

the sequences for detailed analyses using the 3-D reactor dynamics code HEXTRAN 

[1] were considered. 

2 NODALIZATION MODEL 

There are two VVER-440 units at the Loviisa site and Unit 1 was chosen as a 

reference unit. Basically both units are similar but there are some differences like a 

little higher (approx. 4 %) primary coolant mass flow rate in Unit 2. 

The input model included a detailed description of the primary circuit with all six 

loops modeled separately. In Figures 2-1 and 2-2 the nodalization model of pressure 

vessel and one of the six loops is depicted, respectively.

2
Concerning pressure vessel nodalization one important feature can be mentioned. The 

volume in the upper plenum at the hot leg elevation (volume # 5) should locate at the 

hot leg elevation and should have the height of the hot leg pipe diameter. The reason 

for this is when the liquid level in the pressure vessel decreases below the hot leg 

elevation, the flow at the same time changes from two-phase to single-phase steam 

flow. Higher volume was found to have a distinct effect on the results. 

As Figure 2-2 shows, Loviisa has a unique feature since there are two loop seals in the 

primary loop. One is in the hot leg and the other in the cold leg. Also reactor coolant 

pump is different from a typical one. Suction takes place from the side and discharge 

is downwards. 

Figure 2-3 shows the nodalization model of the pressurizer. As one can see pressurizer 

is connected to two hot legs (YA13 and YA14). Pressurizer pressure vessel was 

divided into 15 volume. 

High-pressure injection (HPI), make-up injection, letdown and also primary coolant 

purification systems were modeled as shown in figure 2-4. 

The secondary side included a detailed modeling of each steam generator and steam 

lines to the turbines. Also all the main, emergency and auxiliary emergency feed water 

systems were modeled as shown in Figure 2-5. Secondary side was divided only into 

three stacked volumes on the heat transfer tube area. The height of the bottom volume 

was 0.75 m, the middle volume 0.69 m and the top volume 0.549 m. During the 

accident, however, liquid level in steam generators decreased and the number of steam 

generator secondary side volumes can be considered too small. This is due to the fact 

that RELAP5 tends to overestimate heat transfer in steam generator when liquid level 

decreases. A good heat transfer rate is predicted from primary to secondary side in the 

volumes until practically all the water has boiled off and void fraction in the volume 

becomes 1.0. This kind of behavior results in unrealistic coolant temperature behavior 

in the cold leg. When secondary side volume dries out, heat transfer to the secondary 

side stepwise decreases and cold leg coolant temperature stepwise increases. This 

phenomenon was also seen in these analyses, especially in the cases where steam 

generators liquid level decreased slowly. 

3 ASSUMPTIONS AND ANALYZED CASES 

As a reference case an earlier calculation, which was carried out using 3-D reactor 

dynamics code HEXTRAN, was used. In that case control rod group withdrawal took 

place from 1 % power level assuming 2.2 % reactivity worth of the control rod and the 

initial position of the rod 2 m below the top of the core. This assumption made it 

possible to supply maximum reactivity during withdrawal since, according to the 

Technical Specifications, control rods are not allowed to locate deeper during a hot 

standby situation.

3
In order to get the result of RELAP5/MOD3.1 calculation correspond the reference 

result, some tuning calculations were carried out. During these calculations reactivity 

feedback data was tuned to get approximately equal total reactor power at the end of 

withdrawal as in the reference case. 

In RELAP5/MOD3.1 calculation the reactivity insertion as a result of the control rod 

group withdrawal was modeled, e.g. in the base case, in such a way that reactivity 

increased linearly from 0 to 3.6 $ in 100 s. This was due to the fact that the initial 

position of the control rod was the lowest allowed and withdrawal speed was 2 cm/s. 

The calculation was continued in each case at least up to 30 min. Reactor scram was 

not assumed to take place from any reactor protection signal during this period. 

However, it is realistic to assume that after some time (30 min) the operators are able 

to insert the control rods into the core e.g. by disconnecting the electricity from them. 

Then at least the rod, which was withdrawn, drops down into the core because it could 

not be mechanically stuck. 

In one case analysis was continued up to 60 min without reactor scram since the 

accident had developed to a boiling-condensing mode and the length of the mode was 

examined. After the reactor trip starting of the boiling-condensing mode is not very 

probable since there is not any leakage in the primary circuit. When the control rods 

are in the core the reactor can also tolerate more dilution than in the case when the 

control rods are above the core. 

To maximize decay heat power initial power history was assumed to be such that fuel 

had been in the reactor for two years in average. About two days earlier the reactor 

had been scrammed and the operators were returning it back to the power so that the 

reactor had been 15 min at 1 % power level. This kind of power history results to the 

state where both fission power and decay heat power are 0.5 % of the nominal full 

power. On the other hand, to get the weakest reactivity feedback and thus to maximize 

boiling in the core, BOC reactivity feedback was assumed. Reactivity dependency 

from coolant density, fuel temperature and coolant boron concentration is presented in 

the Appendix in RELAP5 input format. 

In the analysis it was assumed that one out two emergency feedwater pumps is 

available and secondary side pressure was approximately at the opening pressure of 

the turbine bypass valves (4.7 MPa). It was also assumed that one out of two high 

head boron injection pumps is available and both HPI pumps from one redundancy. 

These assumption were not, however, considered to have any greater significance to 

the results since the actuation signal was not expected to turn on, at least not in the 

early phase of the accident. Moreover, due to the shut-off head of the HPI pumps, the 

HPI system is capable to inject ECC water into primary circuit only when pressure is 

less than 12.5 MPa. All other ECC and normal operating systems were assumed to 

operate as designed including the letdown system. 

The fuel gap width was adjusted to get the mean fuel temperature of 530 °C as in 

HEXTRAN run. Initial boron concentration was assumed to be 1646 ppm. Main 

primary and secondary side parameter values are presented in Table 2-1.

4
The effect of reactivity worth was studied in such a way that the withdrawal started from different elevations. Altogether six different variations were analyzed. The initial control rod position and reactivity worth in each case is presented in Table 2-2. The difference between case 1 and 2 was due to the fact that the case 2 was calculated in order to find out the effect of unrealistic high (4 $) control rod reactivity worth. When withdrawal begins from 2 m, 3.6 $ reactivity worth can be considered a conservative 

value. 

In addition to 6 cases presented in Table 2-2, two parameter variations were carried out for the case 6 studying the effect of secondary side pressure and feedwater 

capacity. 

4 ANALYSIS RESULTS 

4.1 Base case, reactivity of withdrawal 3.6 $ 

As already mentioned an event, where control rod is locating 2 m below the top of the core and withdrawal velocity is 2 cm/s, was chosen as a base case. The case was used to check that the reactor power at the end of withdrawal was close to the power calculated using HEXTRAN. In Figures 4.1-1...4.1-12 the behavior of the main 

parameters during calculation is presented. 

As a result of the control rod group withdrawal reactor power begins to increase reaching the maximum power of 2342 MW at the point of time when control rod movement stops (Figure 4.1-1). After that feedback from coolant density or core void fraction (Figure 4.1-4) makes core power to decrease first slowly but after 140 s faster, because as a result of low level (< 1.6 m) in steam generators RCP stops in the corresponding loop. In this case all RCPs stop at the same time since in the early phase of the accident the steam generator liquid level behaves in the same way in each steam generator (Figure 4.1-9). Also primary pressure increases in the early phase reaching maximum 16.6 MPa at 205 s (Figure 4.1-7). Thereafter two-phase natural circulation phase begins. Hot leg void fraction maintains at approximately 0.6 (Figure 4.1-5) and the mass flow rate into the core at about 400 kg/s (Figure 4.1-6). Steam condenses totally in the steam generators and coolant is approximately 10 'C subcooled when entering the core (Figure 4.1-2). From 500 s onward the situation 

remains rather constant till the end of the calculation. 

Boron concentration does not decrease in the core volumes from the initial value (Figure 4.1-10). Instead it slightly increases in the early phase of the accident because 

a lot of steam is discharged through the pressurizer safety valves. At certain point of time during the calculation boron concentration in each cold leg loop seal decreases below the initial value (Figure 4.1-1 1) as a result of steam condensation in the steam generators. Degradation is, however, rather small (< 100 ppm) and therefore it does 

not cause any safety problem. 

In the beginning of the accident reactivity increases very strongly (Figure 4.1-12) reaching the maximum value of 0.58 $ at 20 s. After that point of time feedback, especially from coolant density, becomes effective and reactivity begins to decrease.

5
At the moment when maximum core power is reached reactivity is already less than 

0.1$. 

4.2 Base case, reactivity of withdrawal 4.0 $ 

Since previous case did not result in boiling-condensing mode, it was decided to 

increase reactivity from the withdrawal up to 4.0 $. This value can be considered very 

conservative because the initial position of the control rods and the withdrawal time 

was maintained the same. 

The increased reactivity results, of course, in greater maximum power (2588 MW) at 

100 s (Figure 4.2-1). After that the behavior of power does not essentially differ from 

the power behavior in the previous case. RCPs stop some time earlier (130 s) due to 

faster degradation of the steam generators liquid level (Figure 4.2-9). This also results 

in earlier power degradation due to the reactivity feedback. Just before the calculation 

was stopped fission power fades away and the situation develops to a boiling

condensing mode in which the two-phase natural circulation stops. The analysis was 

not, however, continued since at 30 min the operators can be assumed to be able to get 

at least the withdrawn control rod into the core. 

In spite of the greater maximum power in the early phase of the accident primary 

pressure does not rise as high as in the previous case (Figure 4.2-7). This results from 

the fact that power degradation begins 10 s earlier due to stop of RCPs. 

As in the previous case boron concentration in the core volumes does not decrease 

below the initial value (Figure 4.2-10). In the later phase boron concentration strongly 

increases because boiling in the core. Boron concentration in the cold leg loop seals 

behaves also like in the previous case and not any low boron concentration water 

plugs are predicted to form (Figure 4.2-11). 

Also the behavior of reactivity is similar if the last 100 s is not taken into account 

which indicates the beginning of the boiling-condensing mode. The behavior of void 

fraction in the core (Figure 4.2-4) and in hot legs (Figure 4.2-5) as well as the core 

inlet mass flow rate (Figure 4.2-6) at the end of calculation also indicates the 

beginning of the boiling-condensing mode. 

As a conclusion it can be mentioned that the increase of reactivity worth from 3.6 $ to 

4.0 $ has only a rather mild effect on the results during the first 30 min. 

4.3 Reactivity of withdrawal 2.2 $ 

In this case withdrawal reactivity was decreased down to 2.2 $. This meant that the 

initial position of the control rod was 1.22 m below the top of the core and withdrawal 

time 61 s. 

When less reactivity is inserted the reactor maximum power does not rise as high as in 

the previous cases; only up to 1707 MW (Figure 4.3-1). Smaller reactor power also 

causes slower degradation of the steam generators liquid level (Figure 4.3-9) and later 

stop (175 s) of RCPs. At about 800 s boiling in the core begins continuing almost 

continuously to the end of the calculation. This results in strong reactivity feedback 

(Figure 4.3-12) and at about 15 min fission power fades away. Steam is not, however,

6
able to flow into the steam generators because of water in the hot leg loop seals blocks 

the steam flow. At 20 min steam pushes the hot leg loop seals open and flows into the 

steam generators condensing there. As a result primary pressure decreases rapidly from 13.6 MPa down to 12.2 MPa (Figure 4.3-7). This makes two-phase natural 

circulation to restart but only for a short period (Figure 4.3-6). When the hot leg loop seals fill with water the earlier boiling mode is restored. 

Boron concentration in the core volumes behaves as in two previous cases; i.e. it does 

not decrease below the initial value during the calculation (Figure 4.3-10). Instead the 

situation in the cold leg loop seals is different. In three loops boron concentration 

decreases below initial concentration (Figure 4.3-11). In YA13 and YA14 loops boron 

concentration decreases less than 100 ppm but in YA 11 loop minimum concentration 

is 300 ppm below initial value. No problem is expected if the control rods are in the 

core but if they are above some kind of power transient can be expected. The power transient depends on the size of diluted water plug, entrance velocity and core 

reactivity level when the water plug enters the core. 

It can be concluded that a continuous boiling-condensing mode was not predicted but 

occasional boiling-condensing mode was developed during the event. This mode is 

characterized by sporadic steam flow through hot leg loop seals and it can be called a 

bubbling mode. 

4.4 Reactivity of withdrawal 1.4 $ 

This case dealt with the event in which initial control rod position was 0.78 m below 

the top of the core. Withdrawal time was 39 s and reactivity worth 1.4 $. 

As a result of withdrawal maximum power reaches "only" 1083 MW (Figure 4.4-1). Power behavior shows that this case does not develop to a bubbling mode like the 

previous case. This can also be seen from the behavior of the core entrance mass flow 

rate (Figure 4.4-6), which never decreases as small as in bubbling mode. Also 

pressure vessel liquid level maintains above hot leg elevation during the calculation 

(Figure 4.4-3). 

As for boron concentration the situation also looks good since both in the core 

volumes (Figure 4.4-10) and in the cold leg loop seals (Figure 4.4-11) boron 

concentration never decreases below the initial value. 

1.4 $ reactivity worth as a result of control rod group withdrawal is not enough to 

remove so much coolant from the primary circuit through the safety and relief valves 

that the liquid level in the pressure vessel would decrease below hot leg elevation. The 

event can be considered as a "traditional" ATWS where reactor power finally settles 

down to a certain low level. 

4.5 Reactivity of withdrawal 1.8 $ 

Since in the previous case reactivity effect caused by the withdrawal was too weak, it 

was decided to increase reactivity assuming that the initial location of the control rod 

was 1 m below the top of the core. Thus withdrawal time was 50 s and the reactivity effect 1.8 $ which was 50 % of the base case reactivity.

7
The peak power in this case reaches 1438 MW (Figure 4.5-1). Power behavior shows 

that fission power does not fade away during the calculation, which is a clear 

indication, that two-phase natural circulation continues throughout the analyzed 

period. This can also be seen from the behavior of the pressure vessel liquid level 

(Figure 4.5-3), void fraction in the core (Figure 4.5-4) and the core entrance mass flow 

rate (Figure 4.5-6). 

Boron concentration behavior is also favorable due to the primary circuit behavior. 

Boron concentration in the core volumes (Figure 4.5-10) and in the cold leg loop seal 

(Figure 4.5-11) does not decrease below the initial boron concentration except 

temporarily in the loop YA13 (<50 ppm). 

As a conclusion it can be mentioned that a withdrawal, starting from 1 m depth, is not 

able to decrease primary coolant mass inventory so much that liquid level in the 

pressure vessel would drop below the hot leg elevation and two-phase natural 

circulation would stop. 

4.6 Reactivity of withdrawal 1.9 $ 

Since the previous case did not result in boiling-condensing mode, it was decided still 

increase the withdrawal depth; this time, however, only a little bit from 1 m to 1.06 m. 

This increased reactivity worth from 1.8 $ to 1.9 $. The reason for such a small 

increment was that there was a feeling that the "edge" was quite close. 

It can be easily seen from the reactor power behavior that the situation changes 

drastically compared with the previous one (Figure 4.6-1). When reactor remains on 

decay power it means that two-phase natural circulation stops due to the pressure 

vessel liquid level dropping below hot leg elevation (Figure 4.6-3). This time it was 

also decided to find out how long time it takes until two-phase natural circulation 

restarts and how the reactor behaves during that period. Therefore calculation was 

continued up to 1 h point of time. 

The reactor is under boiling-condensing mode altogether for 35 min (Figure 4.6-6). At 

1500 s two-phase natural circulation restarts for a short period but it is not until at 

3100 s when the two-phase natural circulation restarts and continues at least up to the 

point of time when calculation is stopped. The reason for the second restart is that 

large amount of steam enters the steam generators and condenses there and as a result 

pressurizer level decreases below the HPI pumps set point. At the same time primary 

pressure decreases below the HPI pumps shut-off head and water draining from the 

pressurizer together with high-pressure injection increases the pressure vessel coolant 

mass inventory so much (Figure 4.6-8) that two-phase natural circulation can restart. 

Like in all other cases boron concentration in the core volumes does not decrease 

below the initial value (Figure 4.6-10). Instead in three cold leg loop seals boron 

concentration decreases below the initial concentration after 1500 s (Figure 4.6-11). 

The drop is, however, rather small since minimum concentration in the loop YA16 is 

only approximately 200 ppm below the initial concentration.

8
Reactivity of withdrawal 1.9 $ and turbine bypass not available

In the all calculated cases it was assumed that turbine bypass valves are operating as 

designed. The valves start to open at 4.7 MPa and are fully open at 5.1 MPa. This resulted in situation in which heat transfer to the secondary side was better than in the 

case if only secondary side relief valves would be available. These vales start to open at 5.2 MPa and are fully open at 5.4 MPa. Therefore it was decided to calculate a case 

where only secondary side relief valves were assumed to be available and thus the initial secondary side pressure was raised from 4.7 MPa up to 5.2 MPa. As a result of this assumption AT between primary and secondary side became smaller which 

decreased heat transfer between primary and secondary side. 

Reactor power behavior in this case is very much the same as in the previous case 

(Figure 4.6.1 -1). Maximum power is a little bit smaller and fission power fades away somewhat earlier. The former is due to the fact that reactivity feedback from coolant 

density is slightly stronger and the latter results from earlier stop of two-phase natural 

circulation (Figure 4.6.1-6) since pressure vessel liquid level decreases more rapidly below the hot leg elevation (Figure 4.6.1-3). Core entrance mass flow rate behavior 

shows that like in the previous case two-phase natural circulation restarts at 1500 s. However, now the two-phase natural circulation does not quickly stop again but after 

a short degradation period it seems to start again. The reason for this is that as a result 

of condensation in the steam generators, liquid in the pressurizer is discharged to the 

pressure vessel raising the liquid level (Figure 4.6.1-3). Another reason is that the primary coolant mass inventory does not decrease as much as in the previous case 

(Figure 4.6.1-8). 

Boron concentration behavior in the core volumes (Figure 4.6.1-10) and in the cold 

leg loop seals (Figure 4.6.1-11) is almost similar in both cases. 

Weaker heat transfer does not seem to have any major effect on the results. The 

situation seems even more favorable during the first 30 min if the criterion is the 

generation of the boiling-condensing mode. 

4.6.2 Reactivity of withdrawal 1.9 $ and one main feedwater pump available 

In the second variation one out of four feedwater (FW) pumps was assumed to be 

available in addition to one emergency feedwater (EFW) pump which was available in 

all other cases. 

During the first 500 s reactor power behavior is quite close to the behavior in the two previous cases (Figure 4.6.2-1). Fast fission power degradation as a result of RCPs 

stop begins this time a little bit later because liquid level in the steam generators decreases more slowly (Figure 4.6.2-9). After 500 s power behavior is, however, completely different. This is due to the fact that as a result of better heat transfer to the secondary side primary coolant mass inventory drops clearly less than in the reference 

cases (Figure 4.6.2-8). Therefore pressure vessel liquid level never decreases below the hot leg elevation (Figure 4.6.2-3) and boiling-condensing mode cannot start. Two

phase natural circulation maintains good during the whole calculation (Figure 4.6.2-6) and the system reaches a stationary state where heat generated in the core is removed from the primary coolant by the steam generators. The only disturbing factor is the

9

4.6.1
letdown system, which decreases the primary coolant mass inventory as a result of 

high pressurizer liquid level. If the operators do not do anything the degradation of the 

mass inventory later ends the stationary state. 

As a concluding remark it can be mentioned that one main FW pump during the 

accident improves essentially the situation and the beginning of boiling-condensing 

mode is not predicted during 30 min. 

5 RUN STATISTICS 

The analyses were run on DEC 3000 AXP Model 700 workstation. The input model 

included 587 volume, 680 junctions and 376 heat structures with 1476 mesh points. 

Each calculation was carried out using a maximum time step size of 0.04 s from 0 s to 

300 s. After that the maximum time step size was increased to 0.1 s and it was used 

during rest of the calculation. Run time statistics from 0 s to 300 s and from 300 s to 

1800 s is presented in Table 5-1. The average grind time 

CPU-time / [(number of vol's) <number of time steps)] 

was about 0.33 ms. 

6 CONCLUSIONS 

This report deals with the regulating control rod group withdrawal from an initial 1 % 

power level. The analysis was carried out using RELAP5/MOD3.1 computer code and 

it focused on studying the effect of the initial control rod location to the primary 

coolant boron dilution. 

The weakness of RELAP5/MOD3.1 code in the analysis of reactivity accidents is the 

fact that it only has the point kinetics calculation capability, which does not take into 

account the change in the axial power profile during the control rod group withdrawal. 

One should, however, bear in mind that from the boron dilution point of view this 

shortcoming is not significant. The aim of the analysis was find out situations where 

boiling-condensing mode begins and as a result a diluted water plug may develop in 

the cold leg loop seal. The analysis covers quite many variations of the initial control 

rod location. Thereby the feedback inaccuracy of the coolant density and the fuel 

temperature can be considered to have only minor effect as compared to the broad 

variation of the control rod reactivity worth. 

The results indicate that the continuous boiling-condensing mode does not develop in 

the analyzed cases. In some cases the occasional boiling-condensing or bubbling 

mode was predicted. This kind of situation does not result in a serious boron dilution 

problem, because steam condensation followed by dilution takes place only 

temporarily when steam can flow through the hot leg loop seals and the steam 

generator primary side is filled with water. Between bubbling periods primary coolant 

is getting "power" to be able to bubble again. The maximum dilution in the cold leg 

loop seal was only about 300 ppm, which corresponds approximately 2.3 % reactivity. 

During natural circulation this kind of dilution will not probably cause any serious

10
power transient though the control rods are above the core. Moreover, the diluted 

water plug hardly flows into the core without any mixing on the way. 

It should be pointed out that due to the numerical diffusion, the boron concentration 

behavior in different parts of the primary circuit is not predicted correctly. Therefore, 

more attention should be paid to phenomena like bubbling of the loop seals and fading 

away of the fission power than to the value of the boron concentration. Also from the 

operators point of view almost the only indication of the boiling-condensing mode is 

the information received from the fission power measurement since there is not any 

pressure vessel liquid level measurement in Loviisa. It is a clear indication of boiling

condensing if fission power fades away in case where control rods are above the core 

and no boron is injected into the primary circuit. 

Another important feature is also the initial position of the control rod since the 

reactivity worth was found to be a sensitive parameter. When initial position was 

changed only from 1 m to 1.05 m (1.8 $ - 1.9 $), thermal-hydraulic behavior 

changed considerably. Good two-phase natural circulation ended and the bubbling 

mode began. 

An important finding was also that even if one MFW pump is available, the situation 

is considerably improved because one MFW pump is capable to maintain steam 

generators liquid level except in the early phase of the accident. A good heat transfer 

to the secondary side ensures that the primary coolant mass inventory does not 

decrease so much that the pressure vessel liquid level would drop below the hot leg 

elevation. 

REFERENCE 

[1] R. Kyrki-Rajamaki, HEXTRAN: Three Dimensional Reactor Dynamics Code 

for VVER-reactor Cores. Proceedings of the International Topical Meeting on 

Advances in Mathematics, Computations and Reactor Physics. Pittsburgh, PA, 

USA, 28.4. - 2.5.1991

11
Parameter Initial value 

Pressure [MPaj 

Upper part of pressurizer 12.30 

Hot leg 12.33 

Cold leg 12.69 

Steam generators 4.71 

Steam header 4.71 

Pressure difference [MPaj 

Pressure vessel inlet - exit 0.35 

Across core 0.23 

Across reactor coolant pump 0.46 

Temperature [°C] 

Hot leg 261.1 

Cold leg 260.9 

Mass flow rate [kg/sl 

Downcomer 8709 

Core 7710 

Core bypass 11.5% 

Steam flow 8.9 

Miscellaneous 

Pressurizer liquid level 3.88 m 

Steam generator liquid level 2.22 m 

Core power 13.84 MW 

Steam generators power 16.46 MW 

Average linear power 1.44 W/cm 

Maximum linear power 3.25 W/cm 

Table 2-1 Initial value of the main parameters 

Case Position of the control rod Reactivity worth 

[cm] [$1 

1 200 3.6 

2 200 4.0 

3 122 2.2 

4 78 1.4 

5 100 1.8 

6 105 1.9 

Table 2-2 Initial control rod position and reactivity worth

12
Reactivity CPU time [s] Average time CPU time [s] Average time 

worth [$] 0 - 300 step size [ms] 300 --* 1800 step size [ms] 

3.6 1893 30.33 2913 99.86 

2.2 2149 27.95 3073 99.91 

1.8 2049 28.70 3066 99.97 

1.4 1442 40.00 2976 100.00 

1.9 1969 28.72 2961 99.97 

1.9 Var. 1 2009 29.65 3086 100.00 

1.9 Var. 2 2110 27.25 3006 99.95

Table 5-1 Run time statistics

13
Figure 2-1 Pressure vessel nodalization

Figure 2-2 Loop YA 11 nodalization

14
Figure 2-3 Pressurizer nodalization

YALl YA13 YA14 

TJ20 3 

I._A 1..._2 1/2•,

YA12 YA15 YA1G

Figure 2-4 HPI, make-up, letdown and purification systems nodalization

15
Figure 2-5 Steam generator YB] ], steam line and feedwater injection system 

nodalization

16
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 3.6 $ 

Total and fission power

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.1-1

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 3.6 $ Core entrance, core exit and saturation coolant temperature

0 200 400 600 800 1000 1200 

Time [s]

1400 1600 1800

Figure 4.1-2 

17

2500 

2000 

§ 1500 

0 1000 

500 

0

Total Apowerr --------- Fission ower .. ...................... .................... ...................

360

340

Core entrance 

Core exit 

... ..... , . .. Saturation .... .... ...... ............. .................... ..................... ......................................... .................... i.................. 

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E 

I--

320 

300 

280 

260

mT I

I

.. ........

................ .................... .....................

............... .................... ........... .......... ....................
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 3.6 $ 

Collapsed liquid level above hot leg center line, in core and in pressurizer 

Above hot leg center line 

L .:, .......... .....•.......... ....... ......... Core C r 

Pressurizer 

........... . .......... ............. 

= -i .. ........ ................ .: ...: ': .......... : :- - ..................... i -l.............. ................. ............ ........ 

7 7: n 7: ' =. ====M- =.===t7.-:n== ====7------- ------ ------ ------- 7 -- ------- -- -----

....

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.1-3

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 3.6 $ 

Void fraction in core volumes 

Top - --------- Second ..... ... .. .............. . ............ .... .................... ................. .................... ................... -.......... T h r 

Thr 

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0 200 400 600 800 1000 1200 

Time [s]

1400 1600 1800

Figure 4.1-4 

18

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 4.0 $ 

Collapsed liquid level above hot leg center line, in core and in pressurizer 

Above hot leg center line S........ . ....... . . ................... ...... . . C o re 

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....... ".. .-.... . . .........-... - -..-.. . ................. ..................... . ........... ...... . 

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0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.2-3

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 4.0 $ 

Void fraction in core volumes

0 200 400 600 800 1000 

Time [s]

1200 1400 1600

Figure 4.2-4

24

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4 

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 4.0 $ 

Pressure in hot leg, cold leg and pressurizer

17 

16 

15 

14 

13 

12 

11 

170 

160 

150 

140 

130 

120 

110

.............. .............................. .................... .. .... .... ....... ... ............. ............... .... .................. 

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0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.2-8

26

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.2-7 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 4.0 $ 

Primary coolant mass inventory
C)i S........ ...................... ... i .. .......... O D' ...... ......... K. -.. i c o oco 

• • . .. . .............................. ..... .... 

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 4.0 $ 

Boron concentration in cold leg loop seal horizontal part

2000 

1950 

1900 

r 1850 

co 1800 

= 1750 

o 1700 

M 

C 

2 1650 

1600 

1550 

1500

Figure 4.2-11

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 4.0 $ 

Total reactivity

0.5 

0

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.2-12

28

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s]

.5 

Cu ci,

-0.5 

-1 

-1.5

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 2.2 $ 

Collapsed liquid level above hot leg center line, in core and in pressurizer

0 200 400 600 800 1000 1200 1400 1600 

Time [s]

1800

Figure 4.3-3

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 2.2 $ 

Void fraction in core volumes

0 200 400 600 800 1000 1200 1400 

Time [s]

1600 1800

Figure 4.3-4

30

8

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C, 

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0

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Total and fission power 

I i i-- Fission power 

.. ... .. ... . . .. -- . .

1200 

1000 

800 

600 

400 

200 

0 

360 

340 

320 

300 

280 

260

(D 

0 

0) 

CL 01) 

0. 

E 0-

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.4-2

35

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.4-1 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Core entrance, core exit and saturation coolant temperature

Core entrance 

Core exit 

S .......... Saturation 

.................... .................... .................... ... ................. .................... .... ..................................... 

......... :.............. •.....{............. .................... .................... ..................... .................... •... ................. !..................... ................... 

4... .. J ...1 [-............... ........... ........ .................... "............ ....... ".................... .................... i.................... :.................... -...................
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Collapsed liquid level above hot leg center line, in core and in pressurizer 

-- - - - - - -- - - -- - - - -- - - - . . . -- - - -- - -- -- - - -- - - -

0 200 400 600 800 1000 1200 

Time [s]

1400 1600 1800

Figure 4.4-3 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Void fraction in core volumes

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.4-4 

36

8 

6 

4 -J 

T_

Above hot leg center line 

Core 

Pressurizer

2 V[ .....

0

0.2 

0.15

7 

C 

0 

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0.1

0.05 

0

-----------------------------------------------------------

:: .• •..... ........................ ........................

.............. 

............ .................. . ....... . .......... ............ ..........

I I I I
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Void fraction in hot leg 

Loop YA1l1Y 

i ! ii ; i i i I .. .......... Loop Y YA12 

........................... ........ ' i........ .......... i.................... ..... .................... L o o p Y A 1 4 1 

... .............. i_! . ............ ............ Loop YA14 ].......... Loo YA14 / 

Loop YA1 5 

LoopYA1 6

S....... ................... 0.. 

0 200

,! Y .. .... ... .. .............. .................... . ................ • ;..... ............... T..... ............... ................. .

400 600 800 1000 1200 1400 1600 1800 

Time [s]

Figure 4.4-5

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Core entrance and bypass mass flow rate

Core entrance 

Core byas 

.............. .. ..... •' ... .............. .................... ................. ......... . ..... .................. ............ 

.. .. .. .... . .. .. .. ... .............. i. . ............. . ............. ...... ................. ... ..... .. .... .. .... .... .. ... .. .. 

L .................... ...... .... .. 

S. ............. ... • ......... .................... .................... .................... ...................... .................... .................... ................... 

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.4-6

37

0.1 

0.08

0.06 

.o 

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0.02 

0

800

600

400 

200

0 

0 

0 

U, 

U, 

0

0 

-200

....................

m
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Pressure in hot leg, cold leg and pressurizer

15 

14.5 

14 

13.5 

13 

12.5 

12 

11.5 

11 

170 

160 

150 

140 

130 

120 

110

Cd) 

CZ) 

42

0 200 400 600 800 1000 

Time [s] 

Figure 4.4-8

1200 1400 1600 1800

38

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.4- 7

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.4 $ 

Primary coolant mass inventory 

I p I
(Nj oE

- 0o

[W] 8A81 pflnbil

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 CONTROL ROD REACTIVITY WORTH 1.8 $ Total and fission power 

1 4 0 0 .. ... .. .................. .................... 1400Tiota power Fission ower 

1200 ~ . . 

0. 600 - . 

400 

200 

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 45 51 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ Core entrance, core exit and saturation coolant temperature 360 , • .. i I • Core entr~ance 1000 . .... . .Core exit 

S.......... Saturation 

84 00 ..... .. ................. i .. ................. .................... •..................... .......................................... ! .................... ! .................... .• .................. . 

6 0 .................... .......... ........... .... ... ........ ................ . ........... .......... 

4 02 0 ............ ................... ...... • ..................... ....... .. • ::.............. ........ .... .. .................... 

2 0 0 . . .... ...... ............. ...... .................... ..................... ... . ......... ............... . .................... .................... . ....... 

-8- ---- 

2 0 ----- - -------------- 0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4. 5-1 

360 

300anc Cor 

lCome [s]t 

Fi....re 4atu5-2o

41
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Collapsed liquid level above hot leg center line, in core and in pressurizer i I I I 

..... ..... ................ .. .. .................... ... i ................. .................... ........... .. .... .. ... 

Above hot leg center line 

-------- Core ................... . . . . . . . Prsuie 

.; ........................ : ..... ............... .. ...... .... .. ................ ....... ....... ........... ........ u riz e r ... 

.............. ................... ...... . ....................... ........ ........... . .. ............ ............. - . ............... ............ .. ......... 

---- i

8 

6 

4 

2 

0 

0.3 

0.25 

0.2 

0.15 

0.1 

0.05 

0

7 

C 

0 

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0

... 

0 200 600 4 800.1000 

.. .. ....... ..... .. .. .. ..... ..... ......... ... .. .....

1200 1400 1600 1800

42

Top 

- --------- Second -.-------.. T hird 

.... . ........ Fourth 

Bottom

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.5-3 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Void fraction in core volumes

0 200 400 600 800 1000 

Time [s] 

Figure 4.5-4

0 ..........
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Void fraction in hot leg

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.5-5 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ Core entrance and bypass mass flow rate

0 200 400 600

Core entrance 

Core bypass 

.. .. .";.................. .................... .. ................ .................... • .................... .................... .................... .................... ÷ ................. 

\

\800. 1000 ........... ....... \.................. ............................... ................... 

................. ---------

800 1000 

Time [s] 

Figure 4.5-6

1200 1400 1600

43

0.2 

0.15

7 

0 

Cu 

*- 0.1

Loop YAll -- Loop YA12 S.......... Lo pY 1 

Loop YA14 

............ ......... .. ........................... L o op Y A 15 

Loop YA16 

.. .. . . . .. .......

0.05 

0

800

600

400

200

0 

Cu

0 

-200

1800

. m I I
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Pressure in hot leg, cold leg and pressurizer

15 

14.5 

14 

13.5 

13 

12.5 

12 

11.5 

11 

170 

160 

150 

140 

130 

120 

110

(:L 

1

'-

a

CZ 

0 C• 

0,

0 200 400 600 800 1000 

Time [s] 

Figure 4.5-8

1200 1400 1600 1800

44

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Primary coolant mass inventory 

0 200 400 600 800.1000

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.5- 7
cm \Jce) 11, Lo)( -LO) - LO) 

D) 0) 0) 0Y) a1 

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0 

CO 

0 0 

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[wdd] uoile.JluemoouoD Uog

.......... ...........

S........... ........................ 

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(160 0 

woo 

0, 

10
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Boron concentration in cold leg loop seal horizontal part

... .......... ..... 1.. 

0 200 400 600 800 1000 

Time [s]

1200 1400 1600

Figure 4.5-11

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.8 $ 

Total reactivity

0 200 400 600 800 1000 1200 

Time [s]

1400 1600 1800

Figure 4.5-12 

46

1900

1850 F

"T 1800 

0. 

S1750 

C 1700 C-) 

0 

8 1650 

0 

o 1600 

1550 

1500

I

1800

....! ......................................... i..................... .................... .............. • :: :................. ................. .................... i............ .... 

:: ~ ~~~~~ <: : .. .... t... ..... .. ...... .... . .. .. .............................. 

Loop YA11 ---" . . V A...... .... ............

----------. Loop YA13 ................. Loop YA14 .......... Loop YA15S ......... Loop YA16S

0.6 

0.4 

0.2 

0 

-0.2 

-0.4
0 

LO ~co 

0 

0 

cr, 

CM 

40 E L. 

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $ 

Collapsed liquid level above hot leg center line, in core and in pressurizer 

Above hot leg center line 

8 core 8 i i : .. ........ ...... ":. .. ..i ...... .-.-.-.---

........ Pressurize 

6 .............. .......... . ... 

'-------------------------------------------------

i ........... ................. ..................................... .....i.. ....................... .... ... .. ............................ .• ................ 

.... ........ . .. ..... ........... .... ..... .................. . .................. ...... . .... ............. ........................... .• • : . . 

0 500 1000 1500 2000 2500 3000 35 

Time [s] 

Figure 4.6-3 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $ 

Void fraction in core volumes Top 

0.4 - --------- Second ........ ..... ..... .......... .............. ........... ............... .......... .. ................ ........ ---- -----

.................. F o u rtl -.......... Bottorr 

0 3 " ".......................... .............. ............ ....................... ............................ .................................... 

"0.3 

0.1........................ ................... . ........ 

co 0.2 ----... ... .. 

0.15 ....... . . . ................................ 

.. .... ................ .. .; 1.......... . ..- . . ..... .... ...... 0 '. .... ".AnuýAu ..0.. .........

0 500 1000 150 Time [s] eu•Fuve 

Figure 4. 6-4

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0 0 0 Cf'

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U) C )C ~ 0)ý 

• '"............. ....................... ............... LO z

........... . ...................... _o 0000008 a . 

c:j5 C ,- c'O" c0rc : . •:, • 0)• • ) 0 

zrC it= E 

.. .... . . . . . ... ........... . 0 . . 

o. .............. .. .. ........... ........................................ ..... .......... i ..... ................ . 

O qD • •0 1- J 0O 0 0 

czo 5 oo o 

F- COj C\ 0 0 00 

6 0 00 0o 0 0 [-] uo..oe.. P.OA r- ,. . .

Pi/IlJ elej Moll $sEAI
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $ 

Pressure in hot leg, cold leg and pressurizer 

15 _ _ Hot leg 

S. Cold leg 

Pressurizer 

14 .., ......... ............ i....................... i.................. i .................. .......................... i................... ...................... 

n 14 

1 3 .. . . . . . .. .. .. ... . ... . .. ... . .. .. .. .. .. .. .. . . .. ... .. .. .. . . ... .. .. .. .. . . ... ... . ..ý " -. ..... 

12 ............................................................................. ................ ............ ............ .................. 

10 

0 500 1000 1500 2000 2500 3000 3500 

Time [s] 

Figure 4.6-7 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $ 

Primary coolant mass inventory 

170 

160 

150 15 ... ........... . 

S130 

120 ........ ....... ....... .................................................................................................................................................. 

110 

0 500 1000 1500 2000 2500 3000 3500 

Time [s] 

Figure 4.6-8

50
Boron concentration [ppm] 

0) -4 CD (O 0) c w.r, oD CO 0 0 0 0Q oD CD C> 01 0 0 0D 0

0

CA 

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C,,

C) 

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C,, 

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01 

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1 1 1 

1

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al~ 0 0) 0-J (D CD0 o) 0) 0 C) 0ý C 0D 

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CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Total and fission power ..... ....... .. ............ i .................... i............... ...... S.~~~ 

! i i i ...-- Fission ower 

......................... ...... 

-----------

1400 

1200 

1000 

800 

600 

400 

200 

0

Core entrance 

Core exit 

S .......... Saturation 

0 200. 400...6.....0 1 

... ...... ............. ... .................... ,.................... •..................... .................... .................... .................... .................... .i....•.-..•.........

0 200 400 600 800 1000 

Time Is] 

Figure 4.6.1-2

1200 1400 1600

1800

1800

53

0 200 400 600 800 1000 1200 1400 1600 

Time [s] 

Figure 4.6.1-1 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 Core entrance, core exit and saturation coolant temperature 360 

340 

320 

300 

280 

260

0 a-.

0 

0.. 

E 

M)

I 
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Collapsed liquid level above hot leg center line, in core and in pressurizer

0I 0 0 00 80 10 120 40 160 10

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.6.1-3

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Void fraction in core volumes

-i _ _ 'Top - Second --- ...................... .. .. .. . .. ...: ....................... ............. . .. ................. .............. ........... -.......... T h r Third Fourth .-.......... Bottom

... ...... .. ... .......... .......................................... ................................... ................................ B t o 

S.... .....................

0 200 400 600 800 1000 1200 1400 1600 

Time [s] 

Figure 4.6.1-4

1800

54

S. . .. .. .. .. .. ... . .o- .:* .. . ..... .. .. .. . . . . . .., .... . . .................................. .

. Above hot leg center line 

Core ...... Pressurizer

8 

6 

4 

2 

0

I.

0.4 

0.35 

0.3 

0.25 

0.2 

0.15 

0.1 

0.05 

0

_J

C 

0 

.a 0-

[ a

•.':,,. 
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Void fraction in hot leg

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.6.1-5 

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Core entrance and bypass mass flow rate 

Core entrance 

Core bypass 

...... "................. .. • ................. ,• , :" " " - •; : ; .............. .. .................... i.................... i................ /. . [ -.. = .: . 

I- - I - - I

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.6.1-6

55

1 

0.8

Loop YA1 1 

Loop YA12 

Loop YA13 

L o o p Y A 14 .............................. ................ Loop YA15 

Loop YA16

7 

Cu 

0 

t0.6 l0.4 1-

0.2 -

0

800 

600 

400 

200 0) 

c) 

CD

0 

-200

.......... 

.................. 

..........
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Pressure in hot leg, cold leg and pressurizer 

Cl 

...... Presi 

... .. ... -- .. .. . . .. ......

15 

14 

13 

12 

11 

10

1800

Figure 4.6.1-7

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Primary coolant mass inventory

0 200 400 600 800 1000 1200 

Time [s]

1400 1600 1800

Figure 4.6.1-8

56

03. 

a) 

co

0 200 400 600 800 1000 1200 1400 1600 

Time [s]

CZ 

c170 

160 

150 

140 

130 

120 

110
OD

CD 

CD 

co

0 

0 

Nt 

E 

00 

C> Cý It 

CD R

(0 

C) 

C)ý 

C\l 

C) 

CD 

\C; 

CD 

,, E 

co P 

A 

rzý 

Cý C> q* 

CD CD C\j 

CD

cv).O 

0 -tý 0

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i6> W 

A t4 Q) 

E Cý :3 w -6 

> 

0 cr 908 

C-) 

< 8 zjý . < 0 

-Cc 0 

(1) 0 c 02 

cc 0 

ý-, m 

no 

2! 

Oo

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0

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1-: U) 

ci C) C) Cý C) 0 C> 0 C> C) 0 C> 0 Cý a 0 0 CV) R - (D 01) 00 t- (0 C\l C\j C\l CM 

[wddl uolluilueoum uojog [w] jaAal p!nbil
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Boron concentration in cold leg loop seal horizontal part

2000 

1900 

1800 

1700 

1600 

1500 

1400 

0 

-1 

-2 

-3 

-4 

-5 

-6

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 1 

Total reactivity 

.............. .................... .......... ............................... 8 0 0 .................... 1 0.......... .................... .................. .......... 

.. ......... .... .. .... ....... ..........I.. . ...... ...... ..... .. ...... .I ........

0 200 400 600 800 1000 1200 

Time [s] 

Figure 4.6.1-12

1400 1600 1800

58

0 200 400 600 800 1000 1200 1400 1600 1800 

Time [s] 

Figure 4.6.1-11
0) 

0) 00

0) 

CD 

CD 

CMJ 

>-

Cco 

co~ 

C) 

0l

C\) 

0cuI 

Lf) .- " 

~.0 

8 

00 

60

0ý

o 0 0 0 0D 0 0) 0 0 C0 0 0) 0 0 0 C'o CD 0 0 D It CM 

[MVIJ .JOMOd

0 0 0 0D 0> 0 

CV) Cf CV) 0 (0M C 

[0.] ajnj~iedwqe

C~j 

6O.0 

w -j 

0-I 

2 cu 

t6> 

'ý00 

z0 Q

C0 

0 

00 

0) 

CD 

0ý 

0 

0 

00 

0 

0 

cli
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 2 

Collapsed liquid level above hot leg center line, in core and in pressurizer 

...... .... .. ....... - . .. ...

............... Above hot leg center line 

Core -. .......... Pressurizer I 

0...... 200...... 400.... 60 800 . 100 1200.................... ..... .......

0 200 400 600 800 1000 1200 

Time [s] 

Figure 4.6.2-3

1400 1600 1800

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 2 

Void fraction in core volumes

0 200 400 600 800 1000 1200 

Time [s]

1400 1600 1800

Figure 4.6.2-4 

60

8 

6 

4 

2 

0

0.4 

0.35 

0.3 

0.25 

0.2 

0.15 

0.1 

0.05 

0

T

C 

0 

0

Top 

S.. Second -. ................................... ..................... ...................... ......................................... ............ . . . . h r Third ............... Fourth .......... Bottom 

.... .................... .... . 

S-----------. . 

....... .. ...... .. ... ... ... ...... .. ... .... ....... ..... ....

I I I

I I I

..................... ................

.. ..... .... .. .... ..
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 2 

Void fraction in hot leg

Loop YAll1 Loop YA1 2 

Loop YA13 S... F i i i .. ........ ... L opYA 3 ........... i.................... i.................... ..................... • .................... ] ............. L o o p Y A 1 4 

S.......... ° p A 5

S..... Loop) YA16 [ .i i ............. ... .... ! ......... ... .............. ..... ... ..............

............... .........

S......... !................. ............ • .......... ......... .................... . .................... .................... .................... .+............... .. ......

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.6.2-5

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 2 Core entrance and bypass mass flow rate 8 0 0 . ........ ..................... ............. . C. . ... .............. .............. 

" 400 

2 0 0 ........ ................... ÷ ..-. }.......... .................. .................... .................... .................... ................... ...................... .................. .. 

0 . ... . ...... 

-200

0 200 400 600 800 1000 

Time [s]

1200 1400 1600 1800

Figure 4.6.2-6 

61

0 1 . ........ ............ 

0.8

7 0 

C

.5

0.6

0.4

0.2 

0-

0 

ca 

M,

. i I
0 

CD 

0-

O 

0O

c~2 & 06.2 

o-*_i 0 " - 0 .................... . ...... ........... ....... .. - . 1.3 Cad 

0 - t 0• -...... ............ .................. .. 

i. ! ................. ... ................... • ... ............. ..... ................... -- ....... ........ o...... LOLu 

c- F->0 0D 80 0 o •o0 0 

DD 0CD1) C) 

0 CO f.. 

< az 

O 00 0 .... ...... ......• .... ......... .... i ... ... .. .... • .................... i .................. .................. O C 0 

0 0 0 

04 ) - CD (•J - 0 0 0 0 0 0 0 0 

0 -~ - ( O C1 CN. j 

---------- J[uo ] io u ~ u S~

NMI/• ejnssOJd [suol] AjoluoAuI SSUl/I
CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 2 

Steam generator liquid level

* ..... : ......... ... ......... ! ...... ................... s .--.................... ... ............ ' ......... ... -----........ ---.................... .. ........................... 

.. ............ .. ... .................... ...... ............ ... .. .. ......... .......... --- ............... ................ ...................... i ..................

Steam generator YB1 1 .... ......... .......................... .................... .................... - S tea m g e nerato r Y B 5 2 .... ...... Steam generator YB13 

Steam generator YB54 

Steam generator YB15 

Steam generator YB56

0 200 400 600 800 1000 

Time [s] 

Figure 4.6.2-9

1200 1400 1600 1800

CRWD (ATWS) ANALYSIS USING RELAP5/MOD3.1 

CONTROL ROD REACTIVITY WORTH 1.9 $, Variation 2 

Boron concentration in core volumes SI I iI 

... .................... ...................... . .. ................ ... ......... ........... --------- Second .. .. .. ...................... ......... ................ Third

Fourth .... Bottom 

. .. ..................... ..................... ...................... i .................. .................... ..................... ..................... .................... .. ... ............. .. . 

..... T.................. .•= ......... -- = 

0 0 0 00 80 10 I I II 1,

0 200 400 600 800 1000 

Time [s] 

Figure 4.6.2-10

1200 1400 1600 1800

63

2.5 

2

1.5 E 

>r

1

0.5 "

0

E 

0 

0. 

Q 

C 

a) 

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2200 

2100 

2000 

1900 

1800 

1700 

1600

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0ý 

00

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C) 

0 

0 

CDJ 

0 

C) 

0~

C'j 

a- 64 

"ý 0ý 

rz 00 E0 

t

0

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[wdd] uopuleJoumO uo.jog

T , I 1 

[$1 ApJOG

CD 

C> 0o

(0 

0, 

0) 

C0 

(D 

Co 

C)
APPENDIX 

Reactor kinetics input

I
* Reactor kinetics 

*- ---- I -------- 2-------- 3 -------- 4 -------- 5 -------- 6-------- 7 

30000000 point table4 

* Reactor kinetics information card 

30000001 gamma-ac 13.8378+6 0.0 274.0 1.0 0.631 

* Fission product decay information 

30000002 ans79-1 

* Power history data 

30000401 1375.0+6 90.0 wk 

30000402 0.0 35.0 hr 

30000403 13.75+6 1000.0 sec 

* Reactivity curve 

30000011 900 

* Kinetics feedback table coordinate data 

*-----1 ---- ---- 2-------- 3 -------- 4 -------- 5 -------- 6-------- 7 

30001911 300.0 400.0 500.0 600.0 650.0 * Density 

30001912 700.0 725.0 750.0 800.0 1000.0

803.0 

0.750

* Kinetics feedback table 

* ----- 1 -------- 2 -------- 3---

30002001 00000000 -6.258 

30002003 00000002 -16.930 

30002005 00000004 -25.895 

30002007 00000100 -8.216 

30002009 00000102 -18.718 

30002011 00000104 -27.540 

30002013 00000200 -9.109 

30002015 00000202 -19.534 

30002017 00000204 -28.291 

30002019 00000300 -9.900 

30002021 00000302 -20.255 

30002023 00000304 -28.955 

30002025 00000400 -11.434 

30002027 00000402 -21.657 

30002029 00000404 -30.244 

30002031 00010000 -6.258 

30002033 00010002 -16.930 

30002035 00010004 -25.895 

30002037 00010100 -8.216 

30002039 00010102 -18.718 

30002041 00010104 -27.540 

30002043 00010200 -9.109 

30002045 00010202 -19.534 

30002047 00010204 -28.291 

30002049 00010300 -9.900

* Moderator T, Feedback = ZERO

900.0 

1.050

1100.0 * Doppler

1.500 * Boron

--- 4------- --------- 5 6 -------- 7 

30002002 00000001 -11.836 

30002004 00000003 -20.697 

30002006 00000005 -29.860 

30002008 00000101 -13.706 

30002010 00000103 -22.425 

30002012 00000105 -31.442 

30002014 00000201 -14.558 

30002016 00000203 -23.213 

30002018 00000205 -32.164 

30002020 00000301 -15.313 

30002022 00000303 -23.911 

30002024 00000305 -32.803 

30002026 00000401 -16.777 

30002028 00000403 -25.265 

30002030 00000405 -34.042 

30002032 00010001 -11.836 

30002034 00010003 -20.697 

30002036 00010005 -29.860 

30002038 00010101 -13.706 

30002040 00010103 -22.425 

30002042 00010105 -31.442 

30002044 00010201 -14.558 

30002046 00010203 -23.213 

30002048 00010205 -32.164 

30002050 00010301 -15.313

2

30001921 473.0 

30001931 500.0 

30001941 0.000 

30001942 1.875

633.0 

700.0 

0.375
30002051 

30002053 

30002055 

30002057 

30002059 

30002061 

30002063 

30002065 

30002067 

30002069 

30002071 

30002073 

30002075 

30002077 

30002079 

30002081 

30002083 

30002085 

30002087 

30002089 

30002091 

30002093 

30002095 

30002097 

30002099 

30002101 

30002103 

30002105 

30002107 

30002109 

30002111 

30002113 

30002115 

30002117 

30002119 

30002121 

30002123 

30002125 

30002127 

30002129 

30002131 

30002133 

30002135 

30002137 

30002139 

30002141 

30002143 

30002145 

30002147 

30002149 

30002151 

30002153 

30002155 

30002157 

30002159 

30002161

00010302 

00010304 

00010400 

00010402 

00010404 

01000000 

01000002 

01000004 

01000100 

01000102 

01000104 

01000200 

01000202 

01000204 

01000300 

01000302 

01000304 

01000400 

01000402 

01000404 

01010000 

01010002 

01010004 

01010100 

01010102 

01010104 

01010200 

01010202 

01010204 

01010300 

01010302 

01010304 

01010400 

01010402 

01010404 

02000000 

02000002 

02000004 

02000100 

02000102 

02000104 

02000200 

02000202 

02000204 

02000300 

02000302 

02000304 

02000400 

02000402 

02000404 

02010000 

02010002 

02010004 

02010100 

02010102 

02010104

-20.255 

-28.955 

-11.434 

-21.657 

-30.244 

4.697 

-7.165 

-17.140 

2.932 

-8.778 

-18.625 

2.127 

-9.514 

-19.302 

1.414 

-10.165 

-19.902 

0.031 

-11.429 

-21.065 

4.697 

-7.165 

-17.140 

2.932 

-8.778 

-18.625 

2.127 

-9.514 

-19.302 

1.414 

-10.165 

-19.902 

0.031 

-11.429 

-21.065 

12.932 

0.170 

-10.571 

11.323 

-1.300 

-11.925 

10.590 

-1.971 

-12.543 

9.941 

-2.564 

-13.089 

8.681 

-3.716 

-14.150 

12.932 

0.170 

-10.571 

11.323 

-1.300 

-11.925

30002052 

30002054 

30002056 

30002058 

30002060 

30002062 

30002064 

30002066 

30002068 

30002070 

30002072 

30002074 

30002076 

30002078 

30002080 

30002082 

30002084 

30002086 

30002088 

30002090 

30002092 

30002094 

30002096 

30002098 

30002100 

30002102 

30002104 

30002106 

30002108 

30002110 

30002112 

30002114 

30002116 

30002118 

30002120 

30002122 

30002124 

30002126 

30002128 

30002130 

30002132 

30002134 

30002136 

30002138 

30002140 

30002142 

30002144 

30002146 

30002148 

30002150 

30002152 

30002154 

30002156 

30002158 

30002160 

30002162

00010303 

00010305 

00010401 

00010403 

00010405 

01000001 

01000003 

01000005 

01000101 

01000103 

01000105 

01000201 

01000203 

01000205 

01000301 

01000303 

01000305 

01000401 

01000403 

01000405 

01010001 

01010003 

01010005 

01010101 

01010103 

01010105 

01010201 

01010203 

01010205 

01010301 

01010303 

01010305 

01010401 

01010403 

01010405 

02000001 

02000003 

02000005 

02000101 

02000103 

02000105 

02000201 

02000203 

02000205 

02000301 

02000303 

02000305 

02000401 

02000403 

02000405 

02010001 

02010003 

02010005 

02010101 

02010103 

02010105

-23.911 

-32.803 

-16.777 

-25.265 

-34.042 

-1.502 

-11.355 

-21.555 

-3.187 

-12.914 

-22.983 

-3.957 

-13.626 

-23.635 

-4.637 

-14.255 

-24.211 

-5.958 

-15.477 

-25.330 

-1.502 

-11. 355 

-21.555 

-3.187 

-12.914 

-22.983 

-3.957 

-13.626 

-23.635 

-4.637 

-14.255 

-24.211 

-5.958 

-15.477 

-25.330 

6.265 

-4.340 

-15.329 

4.728 

-5.762 

-16.631 

4.028 

-6.410 

-17.226 

3.408 

-6.984 

-17.751 

2.204 

-8.098 

-18.772 

6.265 

-4.340 

-15.329 

4.728 

-5.762 

-16.631

3
30002163 02010200 10.590 30002164 02010201 4.028 

30002165 02010202 -1.971 30002166 02010203 -6.410 

30002167 02010204 -12.543 30002168 02010205 -17.226 

30002169 02010300 9.941 30002170 02010301 3.408 

30002171 02010302 -2.564 30002172 02010303 -6.984 

30002173 02010304 -13.089 30002174 02010305 -17.751 

30002175 02010400 8.681 30002176 02010401 2.204 

30002177 02010402 -3.716 30002178 02010403 -8.098 

30002179 02010404 -14.150 30002180 02010405 -18.772 

30002181 03000000 19.187 30002182 03000001 12.163 

30002183 03000002 5.740 30002184 03000003 0.983 

30002185 03000004 -5.590 30002186 03000005 -10.612 

30002187 03000100 17.709 30002188 03000101 10.751 

30002189 03000102 4.388 30002190 03000103 -0.324 

30002191 03000104 -6.836 30002192 03000105 -11.810 

30002193 03000200 17.035 30002194 03000201 10.107 

30002195 03000202 3.771 30002196 03000203 -0.920 

30002197 03000204 -7.404 30002198 03000205 -12.357 

30002199 03000300 16.438 30002200 03000301 9.537 

30002201 03000302 3.226 30002202 03000303 -1.448 

30002203 03000304 -7.907 30002204 03000305 -12.841 

30002205 03000400 15.280 30002206 03000401 8.431 

30002207 03000402 2.167 30002208 03000403 -2.472 

30002209 03000404 -8.883 30002210 03000405 -13.779 

30002211 03010000 19.187 30002212 03010001 12.163 

30002213 03010002 5.740 30002214 03010003 0.983 

30002215 03010004 -5.590 30002216 03010005 -10.612 

30002217 03010100 17.709 30002218 03010101 10.751 

30002219 03010102 4.388 30002220 03010103 -0.324 

30002221 03010104 -6.836 30002222 03010105 -11.810 

30002223 03010200 17.035 30002224 03010201 10.107 

30002225 03010202 3.771 30002226 03010203 -0.920 

30002227 03010204 -7.404 30002228 03010205 -12.357 

30002229 03010300 16.438 30002230 03010301 9.537 

30002231 03010302 3.226 30002232 03010303 -1.448 

30002233 03010304 -7.907 30002234 03010305 -12.841 

30002235 03010400 15.280 30002236 03010401 8.431 

30002237 03010402 2.167 30002238 03010403 -2.472 

30002239 03010404 -8.883 30002240 03010405 -13.779 

30002241 04000000 21.736 30002242 04000001 14.567 

30002243 04000002 8.008 30002244 04000003 3.151 

30002245 04000004 -3.564 30002246 04000005 -8.694 

30002247 04000100 20.315 30002248 04000101 13.209 

30002249 04000102 6.708 30002250 04000103 1.894 

30002251 04000104 -4.762 30002252 04000105 -9.847 

30002253 04000200 19.667 30002254 04000201 12.590 

30002255 04000202 6.115 30002256 04000203 1.320 

30002257 04000204 -5.308 30002258 04000205 -10.373 

30002259 04000300 19.093 30002260 04000301 12.042 

30002261 04000302 5.591 30002262 04000303 0.813 

30002263 04000304 -5.792 30002264 04000305 -10.838 

30002265 04000400 17.980 30002266 04000401 10.978 

30002267 04000402 4.572 30002268 04000403 -0.172 

30002269 04000404 -6.730 30002270 04000405 -11.741 

30002271 04010000 21.736 30002272 04010001 14.567 

30002273 04010002 8.008 30002274 04010003 3.151

4
30002275 

30002277 

30002279 

30002281 

30002283 

30002285 

30002287 

30002289 

30002291 

30002293 

30002295 

30002297 

30002299 

30002301 

30002303 

30002305 

30002307 

30002309 

30002311 

30002313 

30002315 

30002317 

30002319 

30002321 

30002323 

30002325 

30002327 

30002329 

30002331 

30002333 

30002335 

30002337 

30002339 

30002341 

30002343 

30002345 

30002347 

30002349 

30002351 

30002353 

30002355 

30002357 

30002359 

30002361 

30002363 

30002365 

30002367 

30002369 

30002371 

30002373 

30002375 

30002377 

30002379 

30002381 

30002383 

30002385

04010004 

04010100. 

04010102 

04010104 

04010200 

04010202 

04010204 

04010300 

04010302 

04010304 

04010400 

04010402 

04010404 

05000000 

05000002 

05000004 

05000100 

05000102 

05000104 

05000200 

05000202 

05000204 

05000300 

05000302 

05000304 

05000400 

05000402 

05000404 

05010000 

05010002 

05010004 

05010100 

05010102 

05010104 

05010200 

05010202 

05010204 

05010300 

05010302 

05010304 

05010400 

05010402 

05010404 

06000000 

06000002 

06000004 

06000100 

06000102 

06000104 

06000200 

06000202 

06000204 

06000300 

06000302 

06000304 

06000400

-3.564 

20.315 

6.708 

-4.762 

19.667 

6.115 

-5.308 

19.093 

5.591 

-5.792 

17.980 

4.572 

-6.730 

23.966 

9.992 

-1.793 

22.597 

8.740 

-2.948 

21.973 

8.168 

-3.474 

21.421 

7.663 

-3.940 

20.348 

6.682 

-4.845 

23.966 

9.992 

-1.793 

22.597 

8.740 

-2.948 

21.973 

8.168 

-3.474 

21.421 

7.663 

-3.940 

20.348 

6.682 

-4.845 

24.973 

10.888 

-0.994 

23.629 

9.658 

-2.128 

23.016 

9.097 

-2.645 

22.474 

8.601 

-3.103 

21.421

30002276 

30002278 

30002280 

30002282 

30002284 

30002286 

30002288 

30002290 

30002292 

30002294 

30002296 

30002298 

30002300 

30002302 

30002304 

30002306 

30002308 

30002310 

30002312 

30002314 

30002316 

30002318 

30002320 

30002322 

30002324 

30002326 

30002328 

30002330 

30002332 

30002334 

30002336 

30002338 

30002340 

30002342 

30002344 

30002346 

30002348 

30002350 

30002352 

30002354 

30002356 

30002358 

30002360 

30002362 

30002364 

30002366 

30002368 

30002370 

30002372 

30002374 

30002376 

30002378 

30002380 

30002382 

30002384 

30002386

04010005 

04010101 

04010103 

04010105 

04010201 

04010203 

04010205 

04010301 

04010303 

04010305 

04010401 

04010403 

04010405 

05000001 

05000003 

05000005 

05000101 

05000103 

05000105 

05000201 

05000203 

05000205 

05000301 

05000303 

05000305 

05000401 

05000403 

05000405 

05010001 

05010003 

05010005 

05010101 

05010103 

05010105 

05010201 

05010203 

05010205 

05010301 

05010303 

05010305 

05010401 

05010403 

05010405 

06000001 

06000003 

06000005 

06000101 

06000103 

06000105 

06000201 

06000203 

06000205 

06000301 

06000303 

06000305 

06000401

-8.694 

13.209 

1.894 

-9.847 

12.590 

1.320 

-10.373 

12.042 

0.813 

-10.838 

10.978 

-0.172 

-11.741 

16.669 

5.045 

-7.020 

15.361 

3.834 

-8.131 

14.764 

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7
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp.. Rev., 

NRCM 1102, and Addendum Numbers, if any.B 3201. 32a2 BIBLIOGRAPHIC DATA SHEET 

(See instructiors on the reverse) 

2. TITLE AND SUBTITLE NUREG/IA-0180 

Application of RELAP5/MOD3.1 to ATWS Analysis of 

Control Rod Withdrawal From 1% Power Level 3. DATE REPORT PUBLISHED 

MONTH YEAR 

June 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

H. Kantee 

7. PERIOD COVERED (Inclusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (ft NRC, provide Division, Office or Region, U.S. Nuclear Regulatry Commission, and mailing address; if contra cto, 

provide name and mailing address.) 

Fortum Engineering, Ltd. 

P.O. Box 10 

00048 Fortum 

FINLAND

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type Same as aboveý if contractcr provide NRC Division. Office or Region, U.S. Nuclear Regulatory Commission. 

and mailing address.)

Division of System Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words ore ss) 

The RELAP5/MOD3.1 computer code has been applied in the analysis of a regulating control rod group withdrawal from 1% 

power level. The analysis has been carried out as an Anticipated Transient Without Scram (ATWS) event. 

The analysis is related to an extensive inherent boron dilution study which was performed to Loviisa Nuclear Power Plant (NPP) in 

the mid 1990's. The main goal of the analysis was to study if during the accident so called boiling-condensing mode develops and 

thereafter a water plug with low boron concentration is formed in the cold leg loop seal.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the rept.) 13. AVAILABILITY STATEMENT 

RELAP5 unlimited 

ATWS 14. SECURITY CLASSIFICATION 

boron dilution (This Page) 

unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
NUREG/IA- 0180 APPLICATION OF RELAP5/MOD3.1 TO ATWS ANALYSIS OF 

CONTROL ROD WITHDRAWAL FROM 1% POWER LEVEL

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC. 20555-0001

years

JUNE 2000