NUREG/IA-0178

International 

Agreement Report 

Cofrentes NPP (BWR/6) 

ATWS (MSIVC) Analysis 

with TRAC-BF1 

1D vs. Point Kinetics and Containment Response 

Prepared by 

A. Escrivd, M.D. Bovea, G. Verd.i, J. L. Mufioz Cobo, UPV, 

F Castrillo, ID 

Universidad Politecnica de Valencia (UPV) 

Departmento de Ingenieria Quimica y Nuclear 

Camino de Vera 14 

46021 VALENCIA 

SPAIN 

IBRDROLA (ID) 

Hermosilla, 3 

28001 MADRID 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0178

International 

Agreement Report 

Cofrentes NPP (BWR/6) 

ATWS (MSIVC) Analysis 

with TRAC-BF1 

1D vs. Point Kinetics and Containment Response 

Prepared by 

A. EscrivA, M.D. Bovea, G. Verdfl, J. L. Mufioz Cobo, UPV, 

E Castrillo, ID 

Universidad Politecnica de Valencia (UPV) 

Departmento de Ingenieria Quimica y Nuclear 

Camino de Vera 14 

46021 VALENCIA 

SPAIN 

IBRDROLA (ID) 

Hermosilla, 3 

28001 MADRID 

SPAIN 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
Abstract

Anticipated transient without scram (ATWS) are considered as events which might pro

duce a severe accident in a boiling water reactor (BWR). It has been selected the most 

unfavourable of the different scenarios that could lead to an ATWS accident: an inadver

tent closure of the main steam isolation valves (MSIVs). 

To mitigate this accident, a borated water solution is injected. This action is an 

alternative way to shutdown the reactor quickly and effectively. This event has been 

analyzed using the TRAC-BF1 computer program.

I
TABLE OF CONTENTS 

1. Introduction ........................................................... 5 

2. Plant M odel .......................................................... 6 

2.1. C ore M odel ........................................................ 7 

2.2. Safety-Relief Valves .................................................. 8 

2.3. M ain Steam Isolation Valve ........................................... 11 

2.4. Reactor Emergency System ........................................... 11 

2.5. High Pressure Safety Systems .......................................... 12 

2.6. Low Pressure Safety Systems .......................................... 13 

2.7. Containment M odel ................................................. 14 

2.8. B oron M odel ...................................................... 16 

3. Initial C onditions ..................................................... 17 

4. T ransient ............................................................ 17 

4.1. Comparison between Point Kinetics and ID Nodal Model ..................... 19 

4.2. Boron Evolution in the Core ........................................... 20 

5. C onclusions .......................................................... 21

2
FIGURES

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE 

FIGURE

1: Model of Cofrentes N.P.P ......................... 

2: Core M odel ................................... 

3: SRV s M odel .................................. 

4: Control Block for Valve Groups I and 2 ............. 

5: Control Block for Valve Groups 3 and 4 ............. 

6: Control Block for Valve Group 5 ................... 

7: HPCS and Boron Injection Model .................. 

8: Containment M odel ............................. 

9: Core Power (Point Kinetics Model) ................. 

10: Core Power (ID Kinetics Model) .................. 

11: Dome Pressure ............................... 

12: Core Flow .................................. 

13: Steam Flow Rate ............................. 

14: Downcomer Level ............................ 

15: Feedwater Flow Rate ........................... 

16: Recirculation Pump Velocity ..................... 

17: Relief Valves Area ............................. 

18: HPCS Injection Velocity ........................ 

19: RCIC Injection Velocity ........................ 

20: Suppression Pool Temperature ................... 

21: Void Fraction in Peripheral Channels (Chan 60) ....... 

22: Boron Concentration in Peripheral Channels (Chan 60) . 

23: Void Fraction in Middle Channels (Chan 61) ......... 

24: Boron Concentration in Middle Channels (Chan 61) .... 

25: Void Fraction in Center Channels (Chan 62) .........

FIGURE 26: Boron Concentration in Center Channels (Chan 62) ................... 39

3

.6 

.8 

.9 

10 

10 

11 

13 

15 

22 

23 

24 

25 

26 

27 

28 

29 

30 

31 

32 

33 

34 

35 

36 

37 

38
TABLES

Characteristics of the SRVs ........................................... 8 

Setpoints of the SRVs ............................................... 9 

Closure of the M SIV ............................................... 11 

Pressure vs. Flow for HPCS ......................................... 12 

Pressure vs. Flow for LPCS .......................................... 14 

Pressure vs. Flow for LPCI .......................................... 14 

A dditive term s ................................................... 16 

Initial Conditions for ATW S analysis ................................... 17 

Sequence of events ................................................ 19

4

TABLE 1: 

TABLE 2: 

TABLE 3: 

TABLE 4: 

TABLE 5: 

TABLE 6: 

TABLE 7: 

TABLE 8: 

TABLE 9:
1 Introduction

The evaluation of events which might lead to severe damage in nuclear reactors, has been 

object of study recently [1], [2], [3]. Events, in which enough failures are assumed, such 

that the reactor is not shutdown, are broadly known as anticipated transients without 

scram (ATWS). It has been selected the most unfavorable of the different scenarios that 

could lead to anticipated transient without scram: an inadvertent closure of the main 

steam isolation valves (MSIVs). 

For a boiling water reactor (BWR), a large amount of hot steam generated in the 

reactor is discharged into the containment pressure suppression pool through the safety

relief valves (SRVs). Temperature of the suppression pool increases and, as a result, the 

containment may eventually be heated and pressurized. 

An increase in the vessel pressure takes place, this causes the activation of the feed

water pump runback and then, the water level in the vessel decreases quickly. To avoid 

this situation two actions are carried out: the recirculation pumps trips and the coolant 

injection through the high pressure safety systems. Water level control accomplished by 

a recirculation pump trip and a proper safety coolant injection can maintain the water 

level at top of active fuel (TAF), introducing negative void reactivity and thus, reducing 

the power production. 

Since these actions are not enough to mitigate the ATWS, at 300 seconds from the 

transient beginning, a borated water solution is injected through the HPCS system line. 

Boron injection at a sufficiently large rate provides an alternative way to quickly and 

effectively shutdown the reactor. The manually actuated standby boron liquid control 

(SBLC) system, injects borated water to reduce reactor power. SBLC system contributes 

to maintain the reactor subcritical. In case of having a severe accident, the failure of the 

borated solution injection must also be included. 

The TRAC-BFI computer program [4] has been used to perform the ATWS event; 

the plant model used in this analysis corresponds to Cofrentes Nuclear Power Plant. The 

scenario assumes that after an MSIV closure an ATWS is initiated, the high pressure safety 

systems are activated and at 300 seconds from the transient, a borated water solution is 

injected. This event has been analyzed considering point kinetics and ID nodal model. 

The purpose of this study is the comparison of TRAC-BF1 results for this type of 

transient, when using point or ID kinetics. The introduction of an specific model for the 

containment is also a contribution of this study. 

In next sections, we will describe briefly the plant model of Cofrentes N.P.P., the 

initial conditions, and the characteristics of the analyzed transient, we will also present 

the results and the main conclusions obtained from the simulation.

5
2 Plant Model 

Cofrentes N.P.P. is a General Electric designed BWR/6 MARK-III plant, with a licensed 

core thermal power of 3015 MWt. The initial plant model used in this analysis, is 

Cofrentes N.P.P. developed for TRAC-BF1, to simulate the transient originated by a 

manual turbine trip, [6] and [7]. 

The main characteristics of the model (see Figure 1), include a 4 ring-11 vessel levels, 

two recirculation loops, with two centrifugal pumps, with a flow control valve (FCV) in 

each loop. Feedwater is supplied by two pumps. One representative steam line supply the 

main turbine with the steam generated in the reactor. This line is equipped with isolation 

valves and 16 safety-relief valves. The core consists of 624 fuel elements distributed in 

three parallel channels, what allows to have a 3-dimensional model. 

J-3 i'1i 

I 0N IN T LI £ RM TURBINE 

,1 I,, '/ ' / , 8 8 8 87 

-7 --- ---- ---- -t-- -- 7.. 

• 'i i -1: - 8Bc ' - ~, ,UPPER, PLENUM / - ' --E 

! !- -i, ..... --- -----... -"-• --] -I -[ - L---- 

frr~- IT LPC1 

J(63 CORE 

:5 

<J-----\ \. :---- - 2 ---~ RECIRCULATION PUMP[ LOWER PLýNOM IPP 

RECIýCULATIO84 

PUMP 

JUIDE G TUBES 

Figure 1: Model of Cofrentes N.P.P.

6
The steady state conditions for the ATWS analyzed are: core power 3015 MW (104.2 

%) and core flow of.8418 kg/s (79 %), which is the minimum flow at licensed power. 

In order to obtain an appropriate plant model for the ATWS conditions, the following 

changes and improvements to the reference model have been performed: 

" First, the CHAN component has been replaced by another one, that models the fuel 

elements corresponding to a full load of GE-il fuel for Cofrentes N.P.P.. 

" The contain MARK III model has been connected to the vessel model. In this way, 

the steam generated in the reactor is discharged into the suppression pool through 

the safety and/or the relief valves (SRVs). 

" The relief valves have been designed as independent valves checked by the control 

system, instead of a multiple bank used in the reference model. 

" The emergency systems have been modeled and connected to the vessel and to the 

containment. 

"* A boron model has been included. 

All the above mentioned changes will be explained with more detail in the following 

sections. 

2.1 Core Model 

The core model considered for Cofrentes N.P.P. corresponds to a full load of GE-i1 fuel 

design. The 624 fuel elements of the core have been distributed in three radial regions: 44 

central, 504 middle and 76 peripheral elements. The power radial distribution considered 

for each of the channels has been the corresponding to a 1.4 peak factor for the fuel ele

ments of the central region, 1.0 for the middle ones and the rest (0.768) for the peripheral 

ones. 

Each region has been modeled with a CHAN component, using 29 axial cells (see 

Figure 2). The upper and lower core cells belong to the inactive zone and in order to 

make consistent our calculations with model used in the 3D simulator [8], we will suppose 

that the active zone is divided into 25 cells of 0.1524 m long, and two cells corresponding 

to the reflector with the same length. The losses of the spacers have been associated to 

the shape losses of the closest section.

7
E

E 

E 

E 

E 

E

V 29 inactive 0.2076 

V 28 reflector 0.1524 

V 27 

V26 

V 25 

V 24 

v 23 

V 22 

v 21 

v 20 

v 19 

V 18 

v 17 

v 16 active 

zone 25 x 0.1524 V 15 

v 14 

v 13 

v 12 

v 1 1 

v 10 

v 

V8 

V7 

v6 

v5 

v4 

V3 

v 2 reflector 0.1524 

I• _ I _ inactive 0.1876

Figure 2: Core Model 

To obtain the core characteristics we have taken as a reference the core model of the 

9th cycle of Cofrentes N.P.P., [9]. 

2.2 Safety-Relief Valves 

The safety-relief valves modeled in the reference model, were designed with a VALVE 

component using the option of multiple bank safety relief with ADS trip. Each valve 

group has the following characteristics:

Group I No. Valves Area(m2) 

1 1 1.4276E-2 

2 1 1.4276E-2 

3 3 4.2828E-2 

4 4 5.7104E-2 

5 7 -9.9932E-2 

Table 1: Characteristics of the SRVs. 

8

9.47 

LEVEL 6 

4.51 

LEVEL 

E4.96
The connection of the multiple bank to the plant model was performed by means 

of a TEE component to the main steam line, and with a BREAK component to the 

containment. 

In the multiple bank valve option, each valves group opens and closes independently, 

based on its own opening and closing pressure setpoints. This mechanism is correct if we 

only have normal setpoints. But for the ATWS analysis, it is necessary to include the 

low-low setpoints, for groups 1 and 2. These setpoints allow to increase the amount of 

discharged steam, after the first opening and closing. But the low-low setpoints cannot be 

included in the multiple bank model. Table 2, shows the normal and low-low setpoints. 

I Normal Setpoints Low-Low Setpoints 

I Popi (MPa) Pc (MPa) Pop2 (MPa) P, (MPa) 

Group 1 7.9835 7.293 7.5 6.7629 

Group 2 8.0525 7.362 7.7766 6.8319 

Group 3 8.1214 7.362 6.9 

Group 4 8.1214 7.362 

Group 5 8.1214 7.431 

Table 2: Setpoints of the SRVs (Relief Function). 

So, the SRV model is not an adequate one for the ATWS analysis, since the pressure 

drop between the dome and the valve is important for the valves opening, specially in the 

first seconds of the transient. To solve these two problems it has been necessary to design 

the SRVs as five independent valves, where each one represents one valve bank group, as 

it is shown in Figure 3.

I CONTAINMENT

['J35 • J36 •' J37 • J38 • 3 

VALVE30 VALVE3 1 VALVE32 VALVE33 VALVE3 4 

"J3 M J32JJ 33 1MKJ 34 MSIV VESE J BO 8 J84 J8 86J J8 TEE80S TEE84 TEE85 TEE86 TEE87 

Figure 3: SRVs Model. 

The valve function is indicated by the control system. Thus, control blocks have been 

designed in order to be able us to reach the normal and low-low setpoints. Figure 4, shows

9
the control blocks used by groups 1 and 2, which need normal setpoints (Popi, P,) and 

low-low setpoints (P-p2, Pc)

This allows to extract higher steam flow from the vessel, since the setpoints of the 

low-low logic are more relaxed than those of the normal one.

Figure 4: Control Block for Valve Groups 1 and 2.

Figure 5 shows the control blocks for groups 3 and 4, which only use normal setpoints 

(P0 p, Pc); that is, the opening pressure remains constant during the successive cycles.

Figure 5: Control Block for Valve Groups 3 and 4.

10
Finally, Figure 6 shows the operation of the group 5 relief valves. The cycles are 

regulated by the normal logic, but besides the relief function, this group must include the 

possibility of core depressurization (ADS), activated by a level trip (Very Very Low Level, 

LI). So, the implementation of this function differences them from the latter group. The 

ADS system function needs the opening of seven of the relief valves, what is equivalent 

to the opening of the modeled group of valves. Once the Li level trip is activated, this 

group opens completely, allowing the release of the generated steam and therefore, the 

pressure in the vessels decreases quickly. 

CTE 

GREQFNGI 

Figure 6: Control Block for Valve Group 5. 

2.3 Main Steam Isolation Valve 

The ATWS analyzed is initiated by the closure of the main steam isolation valve. 

Closure of the MSIV, begins at time zero and the reactor scram fails. The valve is 

completely closed in three seconds, according to the following time-area table. 

Time Area 

Table 3: Closure of the MSIV. 

2.4 Reactor Emergency Systems 

The modeled emergency systems are classified in: 

* high pressure safety systems 

- high pressure core spray, HPCS 

- reactor core isolation cooling, RCIC

11
e low pressure safety systems

- low pressure coolant injection, LPCI 

- low pressure coolant spray, LPCS 

Each of these systems has been modeled by means of a FILL component, and they 

inject coolant in order to balance the loss of water level in the vessel. The injection flow 

can be taken from the condensate tank or from the suppression pool, [3]. But according to 

the increase of the water level in the suppression pool, let us consider that in the analyzed 

case, the injection flow is taken from the pool, [2]. 

2.5 High Pressure Safety Systems 

The HPCS system injects coolant in the upper plenum, corresponding to the level 8 and 

to the radial region 3 of the plant model. The RCIC system injects coolant through the 

same line as the feedwater. 

When the water level in the downcomer falls below the L2 level (Very Low Level), 

these systems are activated with a delay of 37 seconds. The stop signal is the L8 level 

(High Level). In order to simulate this operation, these systems are controlled by the 

TRAC control system. 

HPCS and RCIC suction water is taken from the suppression pool. But the suppression 

pool temperature increases as a result of the amount of steam discharged through the 

SRVs. The HPCS and RCIC pumps cannot operate under an elevated suppression pool 

temperature. They fail when temperature reaches a fixed value. HPCS system is also 

used to allow the boron injection. 

The injected flow by the HPCS system is a function of core pressure, decreasing when 

the core pressure increases, and vice versa, as it can be seen in Table 4. The RCIC flow 

is constant with the core pressure, and its value is 0.03785 m3 /sec. 

Pressure (MPa) I Flow (m3 /sec) 

1.3792 0.056775 

7.9097 0.079485 

8.1166 0.2781975 

Table 4: Pressure vs. Flow for HPCS. 

An additional function of the HPCS system is to allow the injection of a borated 

water solution through its line. This has been modeled using a TEE component (see 

Figure 7), that allows the coolant injection through one of its branches and the borated 

water through the other one.

12
FILL 47 

BORON 

J 47 

VESSEL J 45 J46 .*...-------..c FILL 46 

TEE 45 

Figure 7: HPCS and Boron Injection Model. 

2.6 Low Pressure Safety Systems 

LPCI and LPCS systems inject coolant in the radial region 3 of the vessel. but the former 

injects in the axial level 6 and the latter in the axial level 8. 

When the vessel level decreases below level Li (Very Very Low Level), these systems 

are activated with a delay of 70 seconds, and they stop working when the vessel level 

exceeds the L8 level (High Level). This operation mode has been modeled using the 

TRAC control system. 

Provided that the injection pressures are 1,653 MPa for the LPCI and 1.928 MPa for 

the LPCS, a previous depressurization is not produced and therefore. low pressure safety 

systems are activated, but coolant is not injected because vessel pressure is never lower 

than the injection pressure. 

The injected flow depends on the pressure, according to Tables 5 and 6.

13
Pressure (MPa) Flow (mi/sec) 

0.1013 0.362705 

0.5150 0.3153951

0.825275 0.2838553 

1.0666 0.2523165 

1.27345 0.2207767 

1.445825 0.1892368 

1.583725 0.157697 

1.68715 0.1261582 

1.790575 0.0946184 

1.82505 0.0630786 

1.929 0.0

Pressure (MPa) Flow (m3/sec) 

0.1013 0.8044208 

0.273675 0.7571019 

0.58395 0.6624643 

0.894225 0.5678264 

1.101075 0.4731887 

1.307925 0.378551 

1.445825 0.2839132 

1.54925 0.1892755 

1.653 0.0

Table 5: Pressure vs. Flow for LPCS. Table 6: Pressure vs. Flow for LPCI. 

2.7 Containment Model 

Another improvement incorporated to the plant model, has been the connection to the 

vessel, of a MARK III containment model developed by UPV and Iberdrola. [10]. 

The calculation with TRAC-BF1 is made using the containment component CONTAN. 

which actually is an independent subprogram based on the CONTEMPT-LT program. 

A compartment simulates a volume or room within the containment, composed of 

a vapor and a liquid region. It is assumed pressure equilibrium, but not temperature 

equilibrium between the two regions. 

The containment (CONTAN) component calculates the time variation of compartment 

pressures, temperatures in the liquid pool region and in the vapor atmosphere region above the pool, mass and energy inventories, heat structure temperature distributions, 

and energy exchange with adjacent compartments. 

Each compartment may have energy transfer between the pool and vapor regions for 

the following models: 

- Pool boiling, it is an instantaneous mass transfer. 

- Evaporation, condensation. During a time step the code does not permit both models. 

This model is time dependent. 

Saturation conditions are presumed at the interface, and the heat transfer between the 

pool and the vapor region is equal to the heat transfer between the interface and the bulk 

vapor mixture. The compartments can be communicated using the 'Passive flow .junction' 

component, which simulates pressure-induced convective flow between two compartments. 

There are three types: 

- Single-phase gas flow between the vapor regions of two compartments. Flow may occur 

in either direction.

14
- Single-phase gas flow in one direction only between the vapor regions of two compart

ments. 

- Single-phase gas flow in one direction only between the vapor region of the donor com

partment and the liquid region of the receiver compartment, when the difference of 

pressure between compartments reaches a prescribed value. 

The BREAK and FILL components may be coupled to the containment as an user 

option, specifying the containment compartment to which they are connected. Also the 

VESSEL may be thermally coupled to the containment as another option of the user. By 

specifying which compartment represents the drywell and the position of the vessel above 

the drywell floor. 

As it has been indicated above, the steam is discharged through the SRVs into the 

suppression pool, and accumulated into the wetwell, (see Figure 8).

G DRYWELL 

® ANNULUS 

SWETWELL 

( OUTSIDE 

A JUNCTION DRYWELL-ANNULUS 

AAA VENTS 

A VACUUM BREAKER

81 

FS21 

8e3 

587,1 

Fe5

Fe-6 

FeT

RPV SHIELD WALL 

PEDESTAL 

WEIR 

DRYWELL-WETWELL WALL 

ORYWeLL.WETWELL TOP SLAB 

ANNULUS-WETWELL SLAB 

SHIELD BUILDING

Figure 8: Containment Model.

15
The main elements in the containment model are: compartments, heat structures, RHR heat exchanger, core emergency system (ECCS), vacuum breaker valves and upper 

pool. 

The steam extracted from the vessel through the SRVs is discharged into the suppres sion pool. Its initial temperature is 305.36 K, and its initial volume is 3144.59 m3. Due to 

the amount of steam discharged, the temperature and water level in the suppression pool increases, and then, a heatup and a pressurization in the containment may occur. But if 

the-water temperature in the suppression pool is over a fixed temperature. the HPCS and 

RCIC pumps may not work properly. To avoid this problem, the RHR system removes 

the heat from the suppression pool, working in a cooling mode. Furthermore, if the heat 

capacity temperature limit (HCTL) for the suppression pool is reached, a depressurization 

through ADS system must be activated. 

2.8 Boron Model 

The boron injection from the SBLC provides an alternative method to shutdown the 

reactor, independent of the control rod system. 

The SBLC is manually actuated at 300 seconds from the beginning of the transient, and the borated water is injected through the HPCS system line (as Figure 7 shows), producing a fast reduction of the reactor power. The stop signal is the same for the 

HPCS one; that is, when the water level in the vessel reaches the high level (L8). The 

injected flow is constant: 5,44 i/s with a boron concentration of 21698 ppm. 

The perturbations due to the incorporation of boron in the system, must be included 

in the model. So, the absorption cross sections must be modified. 

The modification carried out for the point kinetics model has been different from the 

one done for the ID model. For the point kinetics model, boron reactivity coefficients has 

to be calculated following the methodology exposed in [11]. 

The new cross sections corresponding to boron injection in ID model, are obtained 

by adding to the initial ones a new term according to Table 7. This table shows these 

additive terms for different void concentrations and for each neutron group, [12], [13]. It can be demonstrated that the value of those additive terms does not depend on the 

burn-up.

Voids %I Group 1 Group 2 

0 3.836E-7 1.247E-5 

40 2.875E-7 9.667E-6 

70 2.160E-7 7.681E-5 

Table 7: Additive terms.

16
3 Initial Conditions

The initial conditions for the starting steady state for the ATWS analysis. are showed in 

the following table, [14]. 

Core Thermal Power (MW/%) 3015./104.2 

Core Flow (Kg/s/%) 8418.5 

By-pass Core Flow 16% 

Dome Pressure (MPa) 7.378 

Feedwater Flow Rate (Kg/s) 1648 

Feedwater Enthalpy (KJ/Kg) 933.44 

F.W. Runback Pressure (MPa) 8.0 

F.W. Flow Stopping Time (s) 40 

Steam Flow Rate (Kg/s) 1648 

Recirculation Flow Rate (Kg/s) 1224 

Table 8: Initial Conditions for ATWS Analysis. 

To adjust the steady state situation with TRAC-BF1 is a hard task to perform, due 

to the amount of variables and parameters to control. 

Once the new stationary situation has been obtained, a null transient from the steady 

state reached, is run to verify stability of the steady state conditions. 

4 Transient 

Once the steady state with the initial conditions needed to the ATWS analysis has been 

achieved, we will study the transient evolution with the TRAC-BF1 code. 

The transient starts with the closure of the MSIV at 0 seconds. During the first 

seconds, the steam flow arriving to the turbine is decreasing, causing the first feedwater 

pump trip because low flow signal in the main steam line, 549 kg/s. Then. the feedwater 

pump runback is produced, because high pressure in the vessel. When the dome pressure 

reaches 8 MP, this trip is activated, with a delay of 40 seconds. 

The water level in the downcomer decreases. When this level reaches L4. the recircu

lation valve runback is activated and it keeps on a fixed opening value of 12.42%. 

As a result of the closure of the MSIV, the large amount of generated steam in the 

reactor, must be discharged into the suppression pool, through the SRVs. When the dome 

pressure reaches the setpoints corresponding to each valve group, they start opening and 

closing, taking into account the normal setpoints. and once they have closed, the low-low 

setpoints are considered. As the suppression pool temperature increases, heat dumped is 

removed by the operation of the RHR system, in the suppression pool cooling mode. 

No credit is taken in this calculation to the recirculation pump trip on high pressure.

17
The water level in the vessel keeps on decreasing until the low level (L3) is reached. This level activates the recirculation pumps low speed transfer, changing their nominal 

speed from 155 rad/s to 38.87 rad/s. 

The downcomer water level continues decreasing until L2. The recirculation pump trip is produced with a delay of 2 seconds when the L2 level is reached. At the same time, but with a delay of 37 seconds, the coolant injection by the HPCS and RCIC systems 

are activated. These measures introduce negative void reactivity and, thus, reduce power production. 

Despite of these actions, the level reaches the Li signal. From this situation, the low pressure safety systems actuates with a delay of 70 seconds. But for the coolant injection to be made through the LPCI and LPCS systems it is necessary a previous core depressurization which is not produced in this analysis. So, these systems are activated 

but there is not coolant injection. 

To mitigate the transient consequences, the borated water injection is activated by means of the SBLC system, through the HPCS line. This action is carried out, manually at 300 seconds, and causes a fast drop of the reactor power, decreasing dome pressure below setpoints SRVs. So, these valves close, and do not discharge more steam into the suppression pool, and its temperature is stabilized below the heat capacity temperature limit (HCTL). The boron concentration in the reactor is linearly proportional with time, and it is uniformly distributed in the core. The analysis finishes studying the reactor evolution until the downcomer level reaches the L8 high level signal. This transient has 

been analyzed using point kinetics and 1D nodal model.

18
Events Point Kinetics ID Kinetics 

MSIV Closure 0.0 0.0 

Water Level L4 1.520 1.524 

1st Feedwater Pump Trip 1.628 1.628 

SRV (Group 1) Opening 2.146 2.149 

Activation of the Feedwater 2.167 2.170 

Pump Runback 

SRV (Group 2, 3 and 4) Opening 2.244 2.247 

SRV (Group 5) Opening 2.473 2.482 

MSIV Fully Closed 3.0 3.0 

Water Level L3 32.044 31.382 

Low Velocity Transfer 32.5 32.0 

Recirculation Pumps 

Feedwater Pump Runback 42.167 42.170 

SRV (Group 5) Closure 72.5 70.5 

Water Level L2 87.216 87.241 

Recirculation Pumps Trip 89.216 89.241 

Water Level Li 111.563 111.156 

SRV (Group 4) Closure 115.0 119.0 

HPCS and RCIC Injection 124.216 124.241 

SRV (Group 3) Closure 142.5 140.5 

LPCS and LPCI Injection 181.563 181.156 

Boron Manual Injection 300.0 300.0 

SRV (Group 2) Closure 319.5 332.0 

SRV (Group 1) Closure 327.0 341.0 

Water Level L8 881.08 908.59 

Table 9: Sequence of Events.

4.1 Comparison between Point Kinetics and ID Nodal Model 

The described transient has been analyzed using point kinetics and the 1D nodal model 

of the TRAC code. 

To calculate the kinetics parameters needed in the input file of the TRAC/BF1 code, 

data from the 3D simulator, used for core design and core following, have been used. 

Specifically, to obtain the point kinetics parameters the PAPU code [15] has been employed 

and for the 1D kinetics, the KINPAR methodology, [16], [17], has been applied. 

From results showed in Table 9, it can be concluded that the temporal behaviour 

of the events using both kinetics match completely. However, if the main variables are 

studied, some differences between this two kinetic models are obtained. The most striking 

difference was the reactor power evolution since in the analyzed case using the point 

kinetics model the characteristic spurious peaks appear, but they do not have significant

19
effects on the rest of the studied variables. 

If the power profile is studied for different instants in the transient, it is observed that 

during the first 300 seconds, when the injection of borated water solution has not been 

produced, the profile is the same as the initial one. 

However, for a time longer than 300 seconds, the power profile changes drastically. A 

deviation of the profile maxima values takes place, initially to the lower axial levels and 

later to the upper ones. This fact is due to the boron action which is concentrated in the 

higher cells during the - 300-400 seconds range (during this period the void fraction is 

close to unity and therefore boron solution does not run along the channels). At -ý 400 

seconds, boron is delivered homogeneously along the channels and then, the maximum 

profile values move to upper levels of the core. But this fact is not meaningful because at 

this time the power reactor is close to residual power. 

This fact explains the different behaviour of the point kinetics parameters, since it has 

been considered for the whole transient, the initial power profile. But the profile evolves 

in a very different way from the initial one along the time when the transient occurs. 

Actually, point kinetics could be applied if it is divided the transient length in temporal intervals, each of them with its appropriate reactivity coefficients. But in practice this is 

not very easy and consequently a ID kinetics is needed for this kind of transients. 

4.2 Boron Evolution in the Core 

This section is focused to study the evolution of the boron concentration along the different 

core channels: peripheral channels (CHAN60), medium channels (CHAN61) and central 

channels (CHAN62). 

For each of this channels it has been represented the void fraction and the boron 

concentration along the time for the different core axial levels: level 3 and level 27, corresponding to the first and last active core cells, respectively, and to the levels 11 and 

19, corresponding to two intermediate axial cells (see Figures 21 to 26). 

The injection of a borated water solution is made manually at 300 seconds in the level 

8 of the vessel and radial ring 3, corresponding to the HPCS system injection line. Boron 

behaves similarly along the three channels in which the core has been modeled. 

At the beginning, the boron concentration rises slowly in the lower axial levels, while 

in the upper ones the concentration reaches very high levels and keep them constant 

approximately at - 80-100 seconds (depending on the channel) as a function of the void 

fraction values. This behaviour is due, mainly, to the increasing impulse of the existing 

steam in the vessel levels where the borated water flow injection is produced. 

The different behaviour in the boron concentration along the axial levels during the 

seconds following the injection is mainly due to the no distribution of the boron in the 

steam. Therefore, when there is a void fraction decrease in the upper levels and the 

void fraction takes a similar value along the channel, the boron concentration changes 

homogeneously.

20
The void fraction decrease in the upper axial level is produced by the slow dilution of 

the boron in the water, what contributes to reduce the thermal power, and consequently, 

diminishes the steam production. 

Studying the behaviour in each channel during the firsts seconds, sharp oscillations 

in the boron concentration are observed. These oscillations coincide with the void frac

tion ones. In this way, when the void fraction decreases, the dilution of boron is favoured, 

diminishing almost instantaneously the boron concentration in the lower part of the chan

nel. 

This fact indicates clearly that boron is not distributed in the steam and therefore, 

until the void fraction is lower than 0.5, boron can not be distributed homogeneously. 

5 Conclusions 

The Main Steam Isolation Valves (MSIV) closure transient, with no automatic insertion 

of control rods (SCRAM), has been analyzed with the TRAC-BF1 code. 

To model the core neutron dynamics, point-kinetics and 1D-kinetics options have been 

used, and the results for both options show no main differences for the overall behaviour 

of the system. 

A model to calculate the response of the containment has been set up, and consistent 

results have been obtained for the evolution of those parameters that are of primary 

concern for this transient (suppression pool temperature). 

After 150 seconds from the transient initiation, the reactor is isolated and with an 

almost constant power around 15 %. If no action is taken. the energy released by the 

SRVs will increase the suppression pool temperature at a constant rate. Therefore, it is 

necessary to activate the boron injection, which will suppress the core thermal power in 

around 50 seconds and further discharges to the pool are eliminated.

21
CORE POWER 

ATWS-POINT KINETICS 

200 .... .......... I 

e)J 100 A 

NI)

0 100 200 300 400 500 600 700 800 885 

time (sec)

Figure 9: Core Power (Point Kinetics Model)
CORE POWER 

ATWS- ID KINETICS 200 -..... ,......",."i.... .... i.... .... I....,.... i....,.... •....,.... I....,.... I....I... 

0

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 10: Corc Power (ID Kinetics Model)
DOME PRESSURE 

ATWS 

8e+06 

POINT 

7e+06 I D 

6e+06 

N5e+06 

4e+06

3e+06 

2.36e+06................................................... ............ 

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 11: Dome Pressure
CORE FLOW 

ATWS 

100-fi 

POINT 

ID 

06 

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 12: Core Flow
STEAM FLOW RATE 

2.21 e+03 ATWS 

2e+03 POINT 

ID 

I e+03 

0i 

0 0 100. 100 200.30 200 40 5.00.60 70 800 900 300 400 500 600 700 800 900

time (sec)

Figure 13: Steam Flow Rate
DOWNCOMER LEVEL 

ATWS 

POINT 

101D 

TAF 

9. 

8" Jet Pump Suction 

7 

-. 7-. 

4 

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 14: Downcorner Level
FEEDWATER FLOW RATE 

ATWS 

1.64e+03 

POINT 

ID 

cj le+03 

00

ol1 0 200 300.. i 500... I 6010 9i0 0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 15: Feedwater Flow Rate
RECIRCULATION PUMP VELOCITY 

ATWS 

156 .,...................................... 

POIN'r 

ID 

100 

N5 

0

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 16: Recirculation Pumnp Velocity
RELIEF VALVES AREA (SRVs) 

ATWS 

0.228...................................................................I 

POINT 

0.2 

IID 

w • 0.1 

0 ........ .... . ...... ........ I ...................................... 

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 17: Relief Valves Area
HPCS INJECTION VELOCITY 

ATWS 4..97 .... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,.... ,..., ... ,....... •...... f 

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

POINT 

4 ID 

3, 

2. 

0 ... ......................................... I ........ 

0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 18: HPCS Injection Velocity
RCIC INJECTION VELOCITY 

ATWS 

1-.... i.... I.... i.... I.... i.... I.... i.... I.... •.... I.... i.... I.... w.... I.... i.... I.... J.... I 

POINT 

ID 

W 

010 200 300 400 500 600 700 800 900

time (sec)

Figure 19: RCIC Injection Velocity
SUPPRESSION POOL TEMPERATURE 

ATWS 

350 . ... ...... 

POINT 

340- 1) 

co 330" 

320 

310, 

305 100 200 300 400 500 600 700 800 9 0 100 200 300 400 500 600 700 800 900

time (sec)

Figure 20: Suppression Pool Temperature
VOID FRACTION CHAN60 

ATWS-ID .. . . . I. . . . .I' ^ •!. •" '-!: -I '•. '-I-; " -I '. 'u.. .I. . .. .I. . .. .I. . . .. I . . . .. I . .o. .. : 

0.9 .... 

0.8 Level- I I 

"0.7 .... . .... :...::'•: 

.. 

0.76. : Level-19 

c If:.. . 0.65 Levl-27 

Iit 

0.4 , ii III\\ j - J~ ll Il ..

0 100 200 300 400 500 600 700 800 900 

time (sec)

Figure 21: Void Fraction in Peripheral Channels (Chan 60)
BORON CONCENTRATION CHAN60 

ATWS-ID

0 100 200 300 400 500 600 700 800 

time (sec)

Level-3 

Level- Il 

Level- 19 

Level-27

900

Figure 22: Boron Concentration in Peripheral Channels (Chan 6()

wA 

C,,
VOID FRACTION CHAN61 

ATWS-ID 

0.988 .v. -;l . . . . . . . . . . . . ... ....... .I - ' , L ev e l- 3 0.9 .. vl ,,, : IU:: l 

0.8 .. " .Level-I I _-......... . ........ ig L 

0.7 i i 

Level-19 

0.7 ' 

"_ 0 Level-27 . All I' 

C0 .5 1- 1 1 4 " 

pw 0.4

0 100 200 300 400 500 600 700 800 900 

time (sec)

Figure 23: Void Fraction in Middle Channels (Chan 61)
BORON CONCENTRATION CHAN61 

ATWS-ID

400 500 

time (sec)

Level-3 

Level- I I 

Level- 19 

Level-27

900

Figure 24: Boron Concentration in Middle Channels (Chan 6 1)

j7

600

500

400

300

200" 

100-

0

j:1 

* ,

ii

100 200 300

6 600 700

"8I 800 . . . . . . . . . .
VOID FRACTION CHAN62 

1 ATWS-1D . . ' . . ... 5 ? .. - . r . -:: -.• , . . .i . . . , .. . , . . . . , . . 

0.9 . . Level-3 

0.8 . .[ Level-ll 

0.7 ) : : ',:' : Level-19 

0•• , -4011 ;;*t Level-27 

44o0.5 , " 

0.4 

0.3 

0.2 ", '3⁄4;..,k I, . , , 

0.1 . 

0 100.200.... 3 6 .4. .5. .. 0.. 0. 9. 0 10O0 200 300 400 500 600 700 800 900

time (sec)

Figure 25: Void Fraction in Center Channels (Chan 62)
BORON CONCENTRATION CHAN62 

ATWS-1D 

500" iLevel-3 

Level-Il 

400" 

Level- 19 

"I Level-27 

S300

' ," .' .. ' " ": '"; .. " 

I ',,, V, I I) ,- .. 

20000. 

100 1 

0 100 200 300 400 500 600 700 800 900 

time (sec)

Figure 26: Boron Concentration in Center Channels (Chan 62)
References

[1] J.G.M. Anderson, L.B. Claassen, S.S. Dua, J.K. Garrett, Analysis of Anticipated Transients Without Scram in Severe BWR Accidents.(Dec, 1987). 

[2] T.J. Liaw, Chin Pan, Gen-Shun Chen, Jung-Kue Hsiue, Analysis of a Main Steam 

Isolation Valve Closure Anticipated Transient Without Scram in a Boiling Water 

Reactor, Nuclear Technology, Vol 88, (Dec, 1989). 

[3] Min Lee, Ein-Chun Wu, A Lomg- Term MAAP 3. OR Analysis of a Severe Anticipated 

Transient Without Scram Accident in a Boiling Water Reactor, N.5uclear Technology, 

Vol 100, (Oct, 1992). 

[4] TRAC-BF1/MODI: An Advanced Best-Estimate Computer Program for BWR Ac

cident Analysis, NUREG/CR-4391, EGG-2417, (1986). 

[5] RETRAN3: A Program for Transient Thermal - Hydraulic Analysis of Complex Fluid - Flow Systems, EPRI NP-7450, (May, 1994). 

[6] Assessment of the Turbine Trip Transient in Cofrentes NPP With TRAC-BF1 ICSP

CO-TTRIP-T, (Jun, 1991). 

[7] Assesment of the One Feedwater Pump Trip Transient in Cofrentes Nuclear Power 

Plant With TRAC-BF1 NUREG/IA-0068, ICSP-CO-TURFW-T, (Apr, 1992). 

[8] SIMULATE-Ill, Studsvik AB Atomenergy Sweden. Studsvik/SOA-89/04, (1989). 

[9] Modelo del Nzicleo de Cofrentes para RETRAN-3, CC-CONUC-11, CN-302, (feb, 

1995). 

[10] J.L. Mufioz-Cobo, G. Verdi, A. Escrivi, Modelizaci6n de la Contenci6n MARK III 

de la C.N. de Cofrentes, para el C6digo TRAC/BF1 y Validacidn de la Misma, (Mar, 

1994). 

[11] J.L. Mufioz-Cobo, G. Verd6, A. Escrivd, Consistent Generation and Functionaliza

tion of One Dimensional Cross Sections and Point Kinetics Feedback Coefficients in 

TRAC-BF1, in progress, (May, 1996). 

[12] J. Martfn, Efecto del Boro en la Secci6n Eficaz de Absorci6n. Aplicaci6n al ATWS 

de C.N. Cofrentes, Revista SNE, (Mar, 1995). 

[13] Effect of the Boron in the Absorption Cross Sections. Application to Cofrentes Cycle 

9, CC-CONUC-8, CN-29, (Feb, 1994). 

[14] C.N. Cofrentes, Actuacicn de los Andlisis de Transitorios Sin Parada Rdpida del 

Reactor (ATWS) a las Nuevas Condiciones de Disefio y Operacidn, IT/CONUC/006, 

N/REF: CN-402, (Mar,1995).

40
[15] A. Escrivd, Tesis Doctoral: Nuevas Aportaciones al Estudio de la Estabilidad de 

Reactores de Agua en Ebullicidn: Desarrollo de un Modelo Fenomenoldgico no Lineal 

para el Estudio de la Dindmica, Valencia, 1998. 

[16] J.L. Muiioz-Cobo, G. Verd6, C. Pereira, A. Escrivd, J. R6denas, F. Castrillo, J. Serra, 

Consistent Generation and Functionalization of One Dimensional Cross Sections for 

TRAC-BF1, Nuclear Technology, No. 107, (1994). 

[17] Generacicn de Pardmetros Cineticos 1D y Puntuales, a Partir de SIMULATE-Ill, 

para su Utilizacijn en TRAC-BF1, DIQN (UPV), (1994).

41
NRC FORM 336 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp.. Rev., NRCM 1102. and Addendum Numbers, If any.) 3=oC, 112. BIBUOGRAPHIC DATA SHEET 

(See miboc,s ca the mve•s) 

2. TITLE AND SUBTITLE NUREG/IA-0178 

Cofrentes NPP (BWR/6) ATWS (MSIVC) Analysis 3. DATE REPORT PUBLSHED 

with TRAC-BF1 3ER MONTH YEAR 

1D vs. Point Kinetics and Containment Response March 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

A. Esciva, M.D. Bovea, G. Verdu, J.L. Munoz Cobo, UPV, Technical 

F. Castrillo, ID 

7. PERIOD COVERED (indusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (iN NRC, provide Division, Oflice orRegion, U.S. NuclwRegulaboy Commiwokn, and maitng address; f coniacta, 

provide name and mailing add-ess) 

Universidad Politecnica de Valencia (UPV) IBRDROLA (ID) 

Camino de Vera 14 Hermosilla, 3 

46021 VALENCIA 28001 MADRID 

SPAIN SPAIN 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Samne as above', if conbtaclr, provide NRC Division, Office or Region, U.S. Nuciw Regulay Commission, 

and mailing ddre&s.) 

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 r200 words crie•s) 

Anticipated transient without scram (ATWS) are considered as events which might produce a severe accident in a boiling water 

reactor (BWR). It has been selected the most favorable of the different scenarios that could lead to an ATWS accident An 

inadvertent closure of the main steam isolation valves (MSIVs). To mitigate this accident, a borated water solution is injected. This 

action is an alternative way to shutdown the reactor quickly and effectively. This event has been analyzed using the TRAC-BF1 

computer program.

12. KEY WORDS/DESCRIPTORS (List woads or phrases that will asist msaerches in locating the report) 13. AVAILABILITY STATEMENT 

TRAC-BF1 unlimited 

BWR/6 14. SECURfTY CLASSFICATION 

ATWS mri paw) 

MSIVC unclassified 

Kinetics (This Rep"t) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
NUREG/IA-0178 COFRENTES NPP (BWR/6) ATWS (MSIVC) ANALYSIS WITH TRAC-BF1 MARCH 2000

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67