NUREG/IA-0226 

 International 

 Agreement Report

 

Assessment of the Turbine Trip 

Transient in Santa María de 

Garoña Nuclear Power Plant 

with TRACE version 4.16 

Prepared by: 

P. Corcuera, E. Rodriguez, and G. Villescas 

University of Cantabria 

ETSIIT 

Av. De los Castros s/n 

39005 Santander, SPAIN 

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 2010 

 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP) 

 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0226

Assessment of the Turbine Trip Transient in 

Santa María de Garoña Nuclear Power Plant 

with TRACE version 4.16

Manuscript Completed: June 2008 

Date Published: April 2010 

Prepared by: 

P. Corcuera, E. Rodríguez, G. Villescas 

University of Cantabria 

ETSIIT 

Av. De los Castros s/n 

39005 Santander, SPAIN 

Antony Calvo, NRC Project Manager 

Prepared for:

Division of Systems Analysis 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

iii 

ABSTRACT

This report presents the results of the assessment of the TRAC/RELAP Advanced 

Computational Engine (TRACE) code, version 4.16, using the model of the Santa María de 

Garoña Nuclear Power Plant to simulate the transient that occurred on June 23, 1992, when a 

false high-level signal in the moisture separator caused an automatic trip of the main turbine. 

The steady state was adjusted by connecting submodels of portions of the system previously 

tuned to the desired conditions. 

The results show a good agreement with data for all the compared variables. The results of the 

calculations were in reasonable agreement with plant measurements. The simulations were run 

on a Pentium IV 3.4 megahertz under Windows XP with 32 bits executable. 

This report was prepared by the Computer Science and Intelligent Systems Group belonging to 

the Applied Mathematics and Computer Science Department of the University of Cantabria, 

which collaborates in the area of simulation with the company Nuclenor S.A., owner of the 

nuclear power station Santa María de Garoña. The Asociación Espanola de la Industria 

Eléctrica (Electric Industry Association of Spain) and Nuclenor S.A. sponsored this work. 
iv 
v 

FOREWORD

This report represents one of the assessment or application calculations submitted to fulfill the 

bilateral agreement for cooperation in thermal-hydraulic activities between the Consejo de 

Seguridad Nuclear (CSN) and the U.S. Nuclear Regulatory Commission (NRC) in the form of a 

Spanish contribution to the NRC’s Code Assessment and Management Program (CAMP), the 

main purpose of which is to validate the TRAC/RELAP Advanced Computational Engine 

(TRACE) code. 

CSN and the Asociación Española de la Industria Eléctrica (Electric Industry Association of 

Spain), together with some relevant universities, have established a coordinated framework 

(CAMP-Spain) with two main objectives: to fulfill the formal CAMP requirements and to improve 

the quality of the technical support groups that provide services to the Spanish utilities, CSN, 

research centers, and engineering companies. 

The AP-28 Project Coordination Committee has reviewed this report, the contribution of one of 

the Spanish utilities to the above-mentioned CAMP-Spain program, for submission to CSN. 

UNESA 

December 2009 
vi 
vii 

CONTENTS

Page

Abstract .......................................................................................................................................iii 

Foreword...................................................................................................................................... v 

Executive Summary.................................................................................................................. .ix 

Abbreviations ............................................................................................................................. xi

1 Introduction ........................................................................................................................1-1 

2 Plant Description................................................................................................................2-1 

3 Transient Description and Plant Response.....................................................................3-1 

4 Code Input Model Description ..........................................................................................4-1 

4.1 Vessel .........................................................................................................................4-1 

4.2 Fuel Elements .............................................................................................................4-5 

4.3 Recirculation Loops.....................................................................................................4-5

4.4 Jet Pumps ...................................................................................................................4-6 

4.5 Guide Tubes................................................................................................................4-6 

4.6 Main Steamlines..........................................................................................................4-7 

4.7 Core Power .................................................................................................................4-7 

4.8 Feedwater ...................................................................................................................4-8 

4.9 Control Systems and Trips..........................................................................................4-8 

5 Steady-State Calculations.................................................................................................5-1 

6 Transient Results and Comparison with Plant Data.......................................................6-1 

7 Run Statistics .....................................................................................................................7-1 

8 Conclusions........................................................................................................................8-1 

9 References..........................................................................................................................9-1 
viii 

Figures

Page

1. Santa María de Garoña NPP functional diagram...........................................................2-3 

2. Reactor power (plant data).............................................................................................3-2 

3. Steam dome pressure (plant data).................................................................................3-2 

4. Reactor level (plant data) ...............................................................................................3-3 

5. Steam flow (plant data) ..................................................................................................3-3 

6. Feedwater flow rate A (plant data) .................................................................................3-4 

7. Feedwater flow rate B (plant data) .................................................................................3-4 

8. Recirculation flow rate A (plant data) .............................................................................3-5 

9. Recirculation flow rate B (plant data) .............................................................................3-5 

10. Core flow (plant data).....................................................................................................3-6 

11. Bypass valve position (plant data) .................................................................................3-6 

12. Santa María de Garoña NPP nodalization .....................................................................4-1 

13. Dimensions of the vessel geometry ...............................................................................4-3 

14. Level distribution in the reactor vessel ...........................................................................4-4 

15. Vessel radial nodalization ..............................................................................................4-6 

16. SNAP nodalization for Santa María de Garoña NPP .....................................................5-2 

17. Steady-state dome pressure ..........................................................................................5-2 

18. Steady-state reactor level ..............................................................................................5-3 

19. Steady-state recirculation flow rate A.............................................................................5-3 

20. Axial void fraction distribution.........................................................................................5-4 

21. Axial pressure distribution ..............................................................................................5-4 

22. Axial liquid temperature distribution ...............................................................................5-5 

23. Axial liquid velocity distribution.......................................................................................5-5 

24. Axial steam velocity distribution .....................................................................................5-6 

25. Steam dome pressure....................................................................................................6-2

26. Reactor level ..................................................................................................................6-2 

27. Core flow ........................................................................................................................6-3 

28. Combined graph of reactor pressure and area of RV/SRV............................................6-3 

29. Feedwater temperature and reactor pressure (real and calculated) ..............................6-4 

30. Total CPU time...............................................................................................................7-1 

31. Timestep size .................................................................................................................7-1 

Tables

Page

1. Initial Plant State ............................................................................................................3-7 

2. Time Sequence for Events in the Turbine Trip Transient...............................................3-8 

3. Components of Garoña NPP Input Deck .......................................................................4-2 

4. Pressure Setting Values and Areas for SRVs................................................................4-8 

5. Reference Steady-State Condition.................................................................................5-1 
ix 

EXECUTIVE SUMMARY

This report presents the results of the assessment of the TRAC/RELAP Advanced 

Computational Engine (TRACE) code, version 4.16, using the Santa María de Garoña Nuclear 

Power Plant (NPP) as a model to simulate the turbine trip transient that occurred there in 

June 1992. 

The Santa María de Garoña NPP is a General Electric boiling-water reactor-3 plant, with a 

nominal core thermal power of 1,381 megawatts thermal, in commercial operation since 1971 

and owned and operated by Nuclenor S.A. 

The objective of this assessment is to generate a Garoña model for TRACE and compare data 

from the model with plant-recorded data during the above-mentioned transient. The model was 

developed with the aid of the Symbolic Nuclear Analysis Package (SNAP) code, version 0.24.1. 

Principal characteristics of the model include a four-ring, 11-axial-level vessel, two recirculation 

loops, and one representative steamline. The control systems and trips were also modeled. 

The SNAP program was used to adjust a reference steady-state condition by connecting 

submodels of portions of the system previously tuned to the desired conditions. The final tuning 

of the input model was done by adjusting the flow area fraction of components in the lower 

plenum of the vessel and the loss coefficients of components near the input of the core. 

As a result of this assessment, a model of the Santa María de Garoña NPP has been developed 

for TRACE that reproduces, in an acceptable manner, the operational transient behavior of the 

plant. Improvement of the recirculation loop model is an area identified for further work. Another 

area of potential improvement is the tuning of the control systems, such as feedwater, pressure, 

and recirculation. 
x 
xi 

ABBREVIATIONS

BWR boiling-water reactor 

CAMP Code Assessment and Management Program 

cm centimeter(s) 

CPU central processing unit 

CSIS-UC Computer Science and Intelligent Systems Group—University of Cantabria 

CSN Consejo de Seguridad Nuclear (Spanish nuclear regulatory commission) 

EPR electric pressure regulator 

FW feedwater 

GE General Electric 

kg kilogram(s) 

l/s liter per second 

m meter(s) 

mm millimeter(s) 

MPa megapascal 

kg/cm^2 kilogram per square centimeter 

oC degrees Celsius 

oK degrees Kelvin 

MSIV main steam isolation valve 

MW megawatt(s) 

MWe megawatt(s) electric 

MWt megawatt(s) thermal 

NPP nuclear power plant 

NRC U.S. Nuclear Regulatory Commission 

Rel pressure relative to the ambient pressure (gauge) 

RV relief valve 

s second(s) 

SNAP Symbolic Nuclear Analysis Package 

SRV safety/relief valve 

SV safety valve 

T/h tonne per hour 

TRACE TRAC/RELAP Advanced Computational Engine 
1-1 

1. INTRODUCTION 

The Computer Science and Intelligent Systems Group of the University of Cantabria (CSIS-UC) 

worked with TRAC-BF1, in 1998–2000, to analyze some transients for the Santa María de 

Garoña Nuclear Power Plant (NPP) and to develop one graphical postprocessing tool for TRAC 

(Ref. 1) as a result of the participation in the CAMP project. In 2007, this group began working 

with the TRAC/RELAP Advanced Computational Engine (TRACE) code. The aim was to obtain 

the model of the Santa María de Garoña NPP and use it to analyze a turbine trip transient that 

occurred in 1992. 

Nuclenor S.A. owns and operates the Santa María de Garoña NPP. This facility has a General 

Electric (GE) boiling-water reactor (BWR)-3, rated at 1,381 megawatts thermal (MWt) and 

connected to the grid in 1971. CSIS-UC has had a close collaboration with Nuclenor in the area 

of simulation with thermal-hydraulic codes. 

The purpose of this report is to document the generation of the model code for TRACE and its 

use to simulate the 1992 turbine trip at the Santa María de Garoña NPP. 

This report consists of the following sections: 

• a brief description of the Santa Maria de Garoña plant, Section 2 

• a description of the plant turbine trip transient, Section 3 

• a description of the model developed for TRACE, Section 4 

• a description of the steady-state calculations, Section 5 

• an analysis and comparison of transient results with plant data, Section 6 

• an analysis of run statistics, Section 7 

The simulations were run on a Pentium 4 workstation, 3 gigahertz under the Windows XP 

Professional 64-bit operating system. 
2-1 

2. PLANT DESCRIPTION 

The Santa María de Garoña NPP is owned by Nuclenor S.A., which is also responsible for its 

operation. 

The plant is a BWR-3, with a Mark I primary containment designed by GE. The plant is rated at 

1,381 MWt. It is located in the province of Burgos, Spain, and was connected to the grid in 

1971. 

The nuclear boiler assembly consists of the reactor pressure vessel and internal reactor 

components, such as the core structure, steam dryer assembly, fuel supports, and control guide 

tubes. The reactor core is made up of 400 fuel assemblies and 97 control rod blades, as well as 

the neutronics instrumentation. At present, it is loaded with GE14 (10x10) elements. However 

the transient that will be compared with a TRACE analysis took place in 1992 (Cycle 17) and the 

reactor core was loaded at that time with the following elements: 

• GE7B (8x8) elements 

• GE8B (8x8) elements 

• GE10 (8x8) elements 

Each control rod blade consists of a sheathed cruciform array of vertical absorber rods made of 

boron carbide. These rods penetrate the core from the bottom. 

The recirculation system provides the hydraulic energy required to force coolant through the 

reactor core, providing it with forced convection cooling. The recirculation system consists 

essentially of two recirculation piping loops located outside the reactor pressure vessel, in the 

drywell area, and includes 20 jet pumps located inside the reactor pressure vessel, between the 

reactor pressure vessel wall and the core shroud. The flow from the recirculation pump is the 

driving force for the jet pump. The primary function of the reactor recirculation system is to 

permit reactor power level changes without changing the position of the reactor control rods. 

Two centrifugal pumps, each driven by an electric motor, supply feedwater. The pumps are 

discharged through spargers located in a ring in the annulus between the core shroud and 

vessel wall. The primary purpose of the feedwater system is to maintain the water level in the 

reactor vessel within a programmed range during all modes of plant operation. In normal 

operation, the level of water in the reactor is controlled by a feedwater controller that receives 

inputs from the reactor vessel water level, steam-mass flow rate, and feedwater-mass flow rate 

transmitters. In turn, the feedwater control system generates signals that regulate the opening 

of the flow control valves. 

The nuclear instrumentation to obtain the necessary information from the local thermal neutron 

flux of the core during full-power reactor operation consists of 22 sets of local power range 

monitors, located radially in the core. Four average power range monitors average signals from 

the 22 local power range monitors to collect information on average power generated in the 

core. 

The main steam system consists of four lines that provide steam to the turbine from the reactor 

vessel. Steamlines run downward, parallel to the vertical axis of the vessel, until they reach the 

elevation at which they emerge from the containment. Two air-operated isolation valves are 

installed on each steamline, one inboard and one outboard of the primary containment 
2-2 

penetration. A flow-restricting nozzle is included in each steamline as an additional engineered 

safeguard to protect against a rapid uncovering of the core in case of a main steamline break. 

Three relief valves (RVs) and three safety/relief valves (SRVs) discharging into the suppression 

pool, and seven safety valves (SVs) discharging into the drywell, are installed on the 

steamlines. The main function of these valves is to protect against overpressure of the reactor 

primary system and the depressurization to allow actuation of low-pressure emergency systems 

in case of a loss-of-coolant accident. 

The primary containment in the Santa María de Garoña NPP is a Mark I. The drywell 

component is a steel “light-bulb shaped” vessel with a spherical lower portion and an upper 

cylindrical portion with the minimum volume necessary to accommodate the reactor vessel and 

ancillary equipment and to allow necessary maintenance and inspection. A bolted head closes 

the top of the cylinder. Reinforced concrete encloses this vessel, providing additional shielding 

and resistance. 

The pressure suppression chamber, or the wetwell, is a toroidal steel vessel that surrounds the 

lower portion of the drywell. Eight circular vent pipes interconnect the wetwell and the drywell. 

The reactor building encloses the containment and also encompasses the refueling area, fuel 

storage facilities, and other auxiliary systems. 

The Santa María de Garoña NPP has the following safeguard systems: 

• isolation condenser system 

• core spray system 

• automatic depressurization system 

• low-pressure cooling injection system 

• high-pressure cooling injection system 

Figure 1 is a functional diagram of the Santa Maria de Garoña NPP obtained from the safety 

parameter display system. 
2-3 

Figure 1 Santa María de Garoña NPP functional diagram
3-1 

3. TRANSIENT DESCRIPTION AND PLANT RESPONSE 

The analyzed transient corresponds to a turbine trip that occurred on June 23, 1992, at 

10:43 pm (22h.43 min). The cause of the trip was the loss of tension in the drain valves to 

moisture condenser separators that created a false high water level signal in the moisture 

separator M1-3A (Ref. 9), when the power plant was operating at 99.5 percent of nominal power 

(1,375 MWt). This signal led to the turbine trip of the plant and the automatic reactor scram. 

The sudden increase of reactor pressure, because of the closure of the stop valves, caused the 

opening of three RVs and two SRVs (one valve didn’t open). Also, the core void fraction content 

was reduced, and the level dropped in a few seconds, reaching the low-level setting 

(+18 centimeters (cm)). Then, the pressure control system opened the bypass valve to the main 

condenser, which allowed it to dominate the pressure. The bypass valve was closed in less than 

30 seconds from the start of the scram. The feedwater control system controlled the reactor 

level by injecting greater flow initially and then the high-level setting (+122 cm) tripped the 

pumps. 

Table 1 summarizes the initial conditions of unity, and Table 2 describes the transient main 

events chronologically. 

The transient’s temporal evolution plots of the most important variables obtained from the 

computer process, some of which have been filtered to eliminate the noise signal plant, are 

shown below: 

• reactor power (Figure 2) 

• steam dome pressure (Figure 3) 

• reactor level (Figure 4) 

• steam flow (Figure 5) 

• feedwater flow rate A (Figure 6) 

• feedwater flow rate B (Figure 7) 

• recirculation flow rate A (Figure 8) 

• recirculation flow rate B (Figure 9) 

• core flow (Figure 10) 

• bypass valve position (Figure 11) 
3-2 

Figure 2 Reactor power (plant data) 

Figure 3 Steam dome pressure (plant data) 

TURBINE TRIP (23-06-1992) 

Reactor power

-10

-5

0

5

10

15

20

25

30

35

40

45

50

55

60

65

70

75

80

85

90

95

100

105

110

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Power (%)

TURBINE TRIP (23-06-1992) 

Reactor Pressure

55

56

57

58

59

60

61

62

63

64

65

66

67

68

69

70

71

72

73

74

75

76

77

78

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Pressure (kg/cm^2)
3-3 

Figure 4 Reactor level (plant data) 

Figure 5 Steam flow (plant data) 

TURBINE TRIP (23-06-1992) 

Reactor Level

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

70

80

90

100

110

120

130

140

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Level (cm)

TURBINE TRIP (23-06-1992) 

Steam flow

-200

0

200

400

600

800

1000

1200

1400

1600

1800

2000

2200

2400

2600

2800

3000

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Flow (T/h)
3-4 

Figure 6 Feedwater flow rate A (plant data) 

Figure 7 Feedwater flow rate B (plant data) 

TURBINE TRIP (23-06-1992) 

Feedwater flow rate A

-200

-100

0

100

200

300

400

500

600

700

800

900

1000

1100

1200

1300

1400

1500

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Flow (T/h)

TURBINE TRIP (23-06-1992) 

Feedwater flow rate B

-200

-100

0

100

200

300

400

500

600

700

800

900

1000

1100

1200

1300

1400

1500

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Tiempo (s)

Caudal (T/h) Flow (T/h)

Time (s)
3-5 

Figure 8 Recirculation flow rate A (plant data) 

Figure 9 Recirculation flow rate B (plant data) 

TURBINE TRIP (23-06-1992) 

Recirculation flow rate A

0

100

200

300

400

500

600

700

800

900

1000

1100

1200

1300

1400

1500

1600

1700

1800

1900

2000

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Tiempo (s)

Caudal (l/s)

TURBINE TRIP (23-06-1992) 

Recirculation flow rate B

0

100

200

300

400

500

600

700

800

900

1000

1100

1200

1300

1400

1500

1600

1700

1800

1900

2000

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Flow (l/s)

Time (s) 

Flow (l/s) 
3-6 

Figure 10 Core flow (plant data) 

Figure 11 Bypass valve position (plant data) 

TURBINE TRIP (23-06-1992) 

Bypass valve position

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Position (%)

TURBINE TRIP (23-06-1992) 

Core flow

5000

6000

7000

8000

9000

10000

11000

12000

13000

14000

15000

16000

17000

18000

19000

20000

21000

22000

23000

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Flow (T/h)
3-7 

Table 1 Initial Plant State 

Parameter Value 

Core thermal power (MWt) 1375 

Generator output (MWe) 445 

Switch mode position RUN 

Reactor dome pressure (kg/cm2) 70.3 

Reactor level RPV (mm) 660 

Core flow rate (T/h) 20.24x103

Feedwater pumps: 

Running 

Selected 

Selected to start by low level 

High-level trip selector 

A and B 

C 

B 

Normal 

Condensate pumps: 

Running 

Selected 

B and C 

C 

Recirculation pumps: 

Running 

Scoop tube blocking by difference signal 

A and B 

Normal 

Control settings: 

Reactor level 

Recirculation pumps 

Turbine/Generator: 

Pressure regulator 

Pressure setting (kg/cm2

) 

Load limiter 

Amplidine 

AUTO/A/3 elements 

Manual 

EPR 

64.7 kg/cm2

90 

Yes 

Local electric distribution NORMAL 
3-8 

Table 2 Time Sequence for Events in the Turbine Trip Transient 

Hour Event State 

22:43:09 Level separator 3A High 

22:43:28 High level separator 

Stop turbine valves close 

General scram A–B 

Bypass valve 

Target Rock Valve C 

Trip 

Trip 

Trip 

Open 

Open 

22:43:29 Isolation Group 2–6 Trip 

22:43:30 RV C 

RV B 

Target Rock Valve A 

RV A 

Target Rock Valve A 

Open 

Open 

Open 

Open 

Closed 

22:43:31 RV A 

Target Rock Valve C 

Closed 

Closed 

22:43:33 RV B Closed 

22:43:34 Manual scram A–B Trip 

22:43:37 Inverse power relay 

Generator block relay 86/G 

RV C 

Trip 

Trip 

Closed 

22:43:46 Level separator 3A Normal 

22:43:57 Bypass valve Closed 

22:45:06 General scram A–B Normal 

22:45:07 Generator block 86/G Normal 

22:45:58 Isolation group 2–6 Normal 
4-1 

4. CODE INPUT MODEL DESCRIPTION 

The development of the TRACE input deck file (Ref. 2, 3, 4, 5, 6) for this analysis was based on 

data taken from previous calculations for TRAC-BF1 code (Ref. 7). Figure 12 shows the main 

features of the Santa María de Garoña NPP model, which includes the reactor vessel and core, 

recirculation loops, and steamlines from the vessel to the turbine valves. 

Figure 12 Santa María de Garoña NPP nodalization 

Table 3 summarizes the components used. The following sections describe the main 

components of the model. 

 

4.1 Vessel

The VESSEL component, which models the reactor vessel, has been divided into 11 axial levels 

and four radial rings. Figure 13 shows the main dimensions of the vessel geometry. 
4-2 

Table 3 Components of Garoña NPP Input Deck 

Component Id. Description 

VESSEL 1 Vessel 

VALVE 3 Recirculation suction valve B 

PUMP 4 Recirculation pump B 

VALVE 5 Recirculation discharge valve B 

PIPE 6 Recirculation discharge line B 

JETP 7 Jet pump loop B 

PIPE 8 Recirculation suction line B 

PIPE 19 Recirculation suction line A 

VALVE 21 Recirculation suction valve A 

PUMP 22 Recirculation pump A 

VALVE 23 Recirculation discharge valve A 

PIPE 24 Recirculation discharge line A 

JETP 25 Jet pump loop A 

TEE 50 Main steamline from the vessel lines 

VALVE 51 RVs d

TEE 52 SVs branch 

VALVE 53 SVs 

VALVE 54 MSIV 

TEE 55 Main steamline to the turbine 

BREAK 56 Wetwell boundary condition 

BREAK 57 Drywell boundary condition 

BREAK 58 Turbine 

BREAK 59 Bypass boundary condition 

PIPE 60 Feedwater line 

FILL 61 Feedwater fill 

VALVE 68 Bypass valve 

VALVE 69 Control valve 

CHAN 71 Hot channels 

CHAN 72 Average channels 

CHAN 73 Peripheral channels 

FILL 82 Guide tube entrance 1 

FILL 83 Guide tube entrance 2 

FILL 84 Guide tube entrance 3 

PIPE 87 Guide tube hot channels 

PIPE 88 Guide tube average channels 

PIPE 89 Guide tube peripheral channels 
4-3 

Figure 13 Dimensions of the vessel geometry 

Each of the 11 axial levels is associated with a component or significant elevation in the vessel; 

the height above each level is defined by the following: 

1. bottom elevation of the control rod drive housings 
4-4 

2. bottom elevation of jet pumps 

3. bottom elevation of suction pipe recirculation loop 

4. core support plate 

5. upper elevation of throat of the jet pumps 

6. upper elevation of upper core grid 

7. upper elevation of feedwater sprays 

8. bottom elevation of the skirt of steam separators 

9. upper elevation of steam separators 

10. upper elevation of steam dryers 

11. bottom elevation of upper dome, assuming that the dome is cylindrical with a radius 

equal to the vessel, and the volume of the dome is the same as the cylinder 

Figure 14 shows the distribution 

chart levels in the reactor vessel. 

Figure 14 Level distribution in the reactor vessel 
4-5 

Four radial rings, as shown in Figure 14, are used to model the vessel, with three of them, the 

internals, to simulate the core, and the fourth as the downcomer. It is assumed that the reactor 

vessel has a cylindrical symmetry, as it uses only one azimuthal segment. The first ring contains 

28 fuel channels; the second ring, 288 fuel channels; and the third ring, 84 fuel channels. 

The perfect separator option is used for axial level 9. The vessel connections to other 

components include the feedwater injection, modelled as a fill governed by the level control 

system, discharging in the downcomer (level 8, ring 4). The steam outlet is located in level 10 

(ring 4). Outlets to recirculation loops from the lower downcomer are located in level 3 (ring 4). 

Recirculation flow mixes in the jet pump with the driven flow from downcomer level 6, to 

discharge into the lower plenum. Channel components representing fuel bundles are connected 

to lower and upper plena, with leakage flow discharging to the bypass flow region (levels 5–7, 

rings 1, 2, and 3). 

4.2 Fuel Elements

The 400 fuel elements (GE-8 type on the date of the transient) of the core were divided into 

three groups, corresponding to the three inner rings of the vessel model. The distribution took 

into account the elements of similar power. Thus, the distribution of the core elements in each 

type of channel is as follows: 

• 28 central high power (hot channel) 

• 288 central average power (average channel) 

• 84 peripheral low power (peripheral channel) 

Each channel was divided into 12 axial nodes, of which 9 are on the active side. Figure 15 

shows the distribution of fuel elements in the core. 

4.3 Recirculation Loops

Both recirculation loops have been modelled, each being divided into five components: 

• PIPE 8-19, representing the suction pipe, from vessel downcomer loop B and A, 

respectively 

• VALVE 3-21, representing recirculation suction valves loop B and A, respectively 

• PUMP 4-22, representing centrifugal pumps loop B and A, respectively 

• VALVE 5-23, representing the recirculation discharge valve loop B and A, respectively 

• PIPE 6-24, representing the discharge pipe up to the jet pumps inlet loop B and A 

respectively 

To simulate the recirculation pumps were used generic characteristic curves of the pumps while 

the moments of inertia were obtained from the data in data sheets of equipment. 
4-6 

4.4 Jet Pumps

Santa María de Garoña NPP has 20 jet pumps, 10 in each recirculation loop. Each jet pump has 

a single nozzle. The 10 jet pumps of one loop have been combined into one single component, 

JETP. The components used in the models are given below. 

• JETP 7, modeling the 10 jet pumps in loop B 

• JETP 25, modeling the 10 jet pumps in loop A 

Figure 15 Vessel radial nodalization 

4.5 Guide Tubes

The 97 guide tubes for the control rods are modelled with three PIPE components, each 

corresponding to an inner ring of the vessel, and the number of guide tubes is a function of the 

number of fuel elements assigned to each ring. The upper guide tube connects to the bottom of 

Hot

Average

Peripheral channel
4-7 

the core bypass. At the entrance of each tube guide, a null FILL is connected to close this 

boundary. 

4.6 Main Steamlines

One single equivalent line has been used to model the four parallel steam pipes. This line 

includes the model of the pressure relief system (3 RVs, 3 SRVs, and 7 SVs), simulated by 

components VALVE. The following components were used for the model of this line from the 

dome of the vessel to the turbine: 

• TEE 50 models the first part of pipes ranging from vessel exit to RVs and SRVs. 

• VALVE 51 models the behaviour of the 3 RVs and 3 SRVs. 

• BREAK 56 represents the pressure boundary condition at the discharge of the RVs and 

SRVs to the wetwell. 

• TEE 52 models the portion of pipes ranging from the RVs and SRVs to the main steam 

isolation valves (MSIVs). 

• VALVE 53 models the behaviour of the 7 SVs. 

• BREAK 57 represents the pressure boundary condition where the SVs discharge to the 

drywell. 

• VALVE 54 models the behaviour of the MSIVs. 

• TEE 55 represents the pipes from the MSIVs to the turbine control/stop valves, and 

branching to the bypass valve. 

• VALVE 68 models the bypass valve. 

• VALVE 69 models the control/stop valves. 

• BREAK 58 represents the pressure boundary condition at the turbine inlet. 

• BREAK 59 represents the pressure boundary condition at the discharge to the main 

condenser. 

Table 4 specifies the opening/closing setting pressure and areas for the valves. 

4.7 Core Power

A reactor point kinetics model with trip-initiated reactivity feedback and trip-initiated scram 

reactivity insertion has been used to calculate the core power rate. The reactivity feedback 

model for void, moderator, and fuel temperature employs reactivity coefficients that have been 

calculated with the PANACEA code by polynomial approximations using core-averaged 

properties. A common axial power distribution is defined for the three types of fuel 

channel/bundles modelled. 
4-8 

Table 4 Pressure Setting Values and Areas for SRVs 

Valve Area (m2) Relative area 

(%) 

Opening setpoint 

pressure (MPa) 

Closing setpoint 

pressure (MPa) 

RV1 0.0058 0.1306 7.277 7.080 

RV2 0.0058 0.1306 7.345 7.149 

RV3 0.0058 0.1306 7.414 7.218 

SRV1 0.009 0.2027 7.683 7.297 

SRV2 0.009 0.2027 7.722 7.336 

SRV3 0.009 0.2027 7.761 7.375 

SV1 0.0145 0.29 8.5 8.1 

SV2 0.0145 0.29 8.6 8.2 

SV3 0.0215 0.43 8.7 8.2 

4.8 Feedwater

The first level extends from the vessel bottom to the top control rod drive housings. The second 

one ends at the jet pumps discharge support ring. The third and fourth ones go from this 

support ring up to the core bottom. 

The feedwater flow is modelled by a constituent FILL, controlled by the feedwater control 

system, injecting water into the vessel downcomer at the temperature defined by the transient 

conditions, through a component PIPE. 

4.9 Control Systems and Trips

The typical BWR control systems have been modelled: level control, pressure control, and 

recirculation control systems. Additionally, trips that represent the reactor protection system and 

other automatic actions have been developed. It is noted that, during the transients analyzed, 

some manual actions were initiated, such as the speed control of a group of recirculation 

pumps. 

 
5-1 

5. STEADY-STATE CALCULATIONS 

A nominal steady state of the system was defined as a reference condition, which was obtained 

from the recorded data before the transient occurred. Table 5 shows the reference conditions 

and the values obtained with the TRACE code for this steady state. 

Table 5 Reference Steady-State Condition 

Parameter Plant Data TRACE 

Thermal power 100% (MW) 1375 1375 

Dome pressure (kg/cm2 rel) 70.26 70.48 

Reactor level (cm rel1) 66.0 66.02 

Recirculation flow (kg/s) 2617.0 2630.02 

Core flow (T/h) 20240 18647.7 

Feedwater flow (kg/s) 677.56 689.10 

Steam flow (kg/s) 685.78 688.7 

Feedwater temperature (K) 452.0 452.8 

To adjust the steady state with TRACE, the approach was to adjust components or submodels 

separately, with the proper boundary conditions, and assemble them one by one to build up the 

entire model. For this purpose, the Symbolic Nuclear Analysis Package (SNAP) program, 

version 0.24.1 (Ref. 8), was used. Partial models were adjusted for recirculation, steamlines, 

channels, vessel, and control systems. The final tuning of the input model was obtained by 

adjusting the flow area fraction of components in the lower plenum of the vessel and the loss 

coefficients of components near the input to the core. Figure 16 shows the full model done with 

SNAP, and Figures 17–19 plot some relevant variables to reach the steady state. 

Some problems were encountered with the nodalization generated with SNAP, because the 

properties of components were incomplete. These situations were resolved by consulting the 

user’s manual and by using the input samples provided with TRACE. 

Table 5 presents the final steady state reached for the most significant variables. It is noted that 

the conditions obtained by TRACE closely approximate the referenced steady state. The 

greatest error occurs in core flow (7.8 percent lower in TRACE). A null transient was run after 

the steady state was reached, to verify the stability of the steady-state conditions. 

Figures 20–24 show the axial distribution of void fraction, pressure, liquid temperature, liquid 

velocity, and steam velocity for hot, average, and peripheral channels in a steady state. 

 

1

 The centimeters relative to zero-scale (12.22 m from the bottom head) correspond to the bottom of the 

steam separators. 
5-2 

Figure 16 SNAP nodalization for Santa María de Garoña NPP 

Figure 17 Steady-state dome pressure 
5-3 

Figure 18 Steady-state reactor level 

Figure 19 Steady-state recirculation flow rate A 
5-4 

Figure 20 Axial void fraction distribution 

Figure 21 Axial pressure distribution 

Void fraction distribution

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 2 4 6 8 10 12

Axial nodes

Void fraction

Hot chan

Average chan

Peripheral chan

Pressure distribution

70.4

70.6

70.8

71

71.2

71.4

71.6

71.8

0 2 4 6 8 10 12

Axial nodes

Pressure (Kg/cm2)

Hot chan

Average chan

Peripheral chan
5-5 

Figure 22 Axial liquid temperature distribution 

Figure 23 Axial liquid velocity distribution 

Liquid temperature distribution

544

546

548

550

552

554

556

558

560

0 2 4 6 8 10 12

Axial nodes

Temperature (oK)

Hot chan

Average chan

Peripheral chan

Liquid velocity distribution

0

1

2

3

4

5

6

7

8

9

10

0 2 4 6 8 10 12 14

Axial nodes

Velocity (m/s)

Hot chan

Average chan

Peripheral chan
5-6 

Figure 24 Axial steam velocity distribution 

Steam velocity distribution 

0

2

4

6

8

10

12

0 2 4 6 8 10 12 14

Axial nodes

Velocity (m/s)

Hot chan

Average chan

Peripheral chan
6-1 

6. TRANSIENT RESULTS AND COMPARISON WITH PLANT DATA 

The transient simulation using the TRACE code considered the following factors that occurred 

at the plant: 

• The SRV B remained closed. 

• The operator manually reduced the speed of the recirculation pumps (Figures 8, 9). 

During this time, the scoop tube of group B was blocking until an automatic run back 

occurred, caused by the feedwater flow falling below 20 percent of the nominal flow 

(Figures 6, 7). 

After a null transient, the transient began with the closing of control valves in 3 seconds. For all 

other active components during the transient, a lookup table was added for the feedwater flow 

injection, speed of recirculation pumps, position of control valve, and core power. It has also 

been introduced by tabular method the feedwater temperature change caused by the loss of 

steam into the shell of heaters in the feedwater system. 

Figures 25–27 show plot comparisons of calculated and measured values for the most relevant 

variables. All plots include 32 seconds of steady state before the initiation of the transient. 

Figure 28 shows a combination plot of the area of RVs and SRVs and the reactor pressure. It 

shows the opening of three RVs and one SRV. Figure 29 shows another combination plot that 

represents the temperature feedwater, the plant data reactor pressure, and the calculated 

reactor pressure. 

These figures show that the calculated variables with TRACE are in acceptable agreement with 

the measured values. For this purpose, it was necessary to adjust the values of the abovementioned tables, especially those relating to temperature feedwater that greatly influence the 

results of the transient. 
6-2 

Figure 25 Steam dome pressure 

Figure 26 Reactor level 

TURBINE TRIP (23-06-1992) 

Reactor level

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

70

80

90

100

110

120

130

140

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Level (cm)

Plant data TRACE

TURBINE TRIP (23-06-1992)

Reactor pressure

55

56

57

58

59

60

61

62

63

64

65

66

67

68

69

70

71

72

73

74

75

76

77

78

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Pressure (kg/cm^2)

Plant data TRACE
6-3 

Figure 27 Core flow 

Figure 28 Combined graph of reactor pressure and area of RV/SRV 

TURBINE TRIP (23-06-1992) 

Core flow

5000

6000

7000

8000

9000

10000

11000

12000

13000

14000

15000

16000

17000

18000

19000

20000

21000

22000

23000

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Flow (T/h)

Plant data TRACE

TURBINE TRIP (23-06-1992)

Pressure and Area RV/SRV

70

71

72

73

74

75

76

77

0 5 10 15

Time (s)

Pressure (kg/cm^2)

0

0.01

0.02

0.03

Area (m^2)

TURBINE TRIP (23-06-1992)
6-4 

Figure 29 Feedwater temperature and reactor pressure (real and calculated) 

DISPARO DE TURBINA (23-06-1992) 

FW Temperature & Pressure

145

150

155

160

165

170

175

180

185

190

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Temperature (oC)

55

56

57

58

59

60

61

62

63

64

65

66

67

68

69

70

71

72

73

74

75

76

77

78

Pressure (kg/cm^2)

Temp FW Pressure Plant Data Pressure TRACE
7-1 

7. RUN STATISTICS 

The simulations were run on a Pentium 4 workstation, 3 gigahertz under Windows XP 

Professional 64-bits operating system. 

Figure 30 is a plot of the total central processing unit (CPU) time, and Figure 31 plots the 

timestep size during the simulation. It shows that the simulated transient runs faster than real 

time, and there is a sudden increase in the CPU time when the transient starts, caused by a 

lower timestep size. Then the timestep size increases, and the total CPU decreases again. 

 

Figure 30 Total CPU time 

Figure 31 Timestep size 

TURBINE TRIP (23-06-1992) 

Total CPU time

0

10

20

30

40

50

60

70

80

90

100

110

120

130

140

150

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

CPU Time (s)

TURBINE TRIP (23-06-1992) 

Timestep size

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180

Time (s)

Timestep size (s)
8-1 

8. CONCLUSIONS 

A model of the Santa María de Garoña NPP for the TRACE code, version 4.16, has been 

developed. The SNAP code, version 0.24.1, was used during the model nodalization and tuning. 

The model has been validated to obtain a steady state and, at the same time, a turbine trip 

transient. In both cases, the calculations were compared with plant data with acceptable results. 

To achieve a steady state, the reference values were the ones used in the model for 

TRAC BF1, with adjustments of a number of parameters of the components regarding the flow 

areas and loss coefficients near the core inlet. 

Core flow is the variable that is responsible for the greatest difference between measurements 

and calculations. Improvement of the recirculation loop model is an area identified for further 

work. Another area of potential improvement is tuning the control systems, such as feedwater, 

pressure, and recirculation. 
9-1 

9. REFERENCES 

1. P. Corcuera, M. Garcés, “Nowadays Tools for Graphical Post-Processing of TRAC-BF1 

Results,” NUREG/IA-0190, U.S. Nuclear Regulatory Commission, Office of Nuclear 

Regulatory Research, 2000. 

2. “TRACE V4.0 User’s Manual,” U.S. Nuclear Regulatory Commission, Office of Nuclear 

Regulatory Research, 2004. 

3. “TRACE V4.16 User’s Manual Volume 1: Input Specification,” Division of Systems 

Analysis and Regulatory Effectiveness, U.S. Nuclear Regulatory Commission, Office of 

Nuclear Regulatory Research, 2004 

4. J.W. Spore, J.S. Elson, S.J. Jolly-Woodruff, T.D. Knight, J.H. Mahaffy, C. Murray, 

“TRAC-M/FORTRAN 90 (Version 3.0), Theory Manual,” Report LA-UR-00-910, Los 

Alamos National Laboratory and Pennsylvania State University, 2000. 

5. J.A. Borkowski, N.L. Wade, et al., “TRAC-BF1/MOD1: An Advanced Best-Estimate 

Computer Program for BWR Accident Analysis,” NUREG/CR-4356, EGG-2626, Idaho 

National Engineering Laboratory, August 1992. 

6. J.A. Borkowski, N.L. Wade, et al., “TRAC-BF1/MOD1 Models and Correlations,” 

NUREG/CR-4391, EGG-2680, Idaho National Engineering Laboratory, August 1992. 

7. “Cuaderno de Cálculo de la Nodalización de la CN de Santa María de Garoña con el 

código TRAC-BF1,” Sección de Sistemas, Nuclenor S.A., 1994. 

8. “Symbolic Nuclear Analysis Package (SNAP), User’s Manual,” Applied Programming 

Technology, Inc., 2007. 

9. “Informe SCRAM por disparo de turbina 23-06-92,” Rev. 1, Sección de apoyo técnico a 

operación, Nuclenor S.A., January 7, 1992.