NUREG/IA-0234

 International 

Agreement Report

Analysis of a Loss of Normal Feedwater

Transient at the Ringhals-3 NPP Using 

RELAP5/Mod3.3 

Prepared by:

J. Bánáti, C. Demaziére, and M. Stålek

Chalmers University of Technology

Department of Nuclear Engineering

S-41296 Gothenburg, SWEDEN

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

July 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-0234

 

Analysis of a Loss of Normal Feedwater 

Transient at the Ringhals-3 NPP Using 

RELAP5/Mod3.3 

Manuscript Completed: May 2010 

Date Published: June 2010 

Prepared by: József Bánáti, Christophe Demazière and Mathias Stålek 

Chalmers University of Technology 

Department of Nuclear Engineering 

S-41296 Gothenburg, SWEDEN 

 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

iii 

ABSTRACT 

This report gives an account on the development and validation of the RELAP5/Mod3.3 model of the 

Ringhals-3 pressurized water reactor against a Loss of Normal Feedwater Transient, which occurred on 

August 16, 2005. The 3rd unit of Ringhals Nuclear Power Plant comprises a 3-loops Westinghouse design 

pressurized water reactor on the Swedish West Coast. 

At first, the RELAP5 model is presented. All the 157 fuel assemblies are modeled individually in the code 

input. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and 

temperature fields between the loops. 

The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. 

Consequently, the turbines were tripped and, due to the low level in the SG-2 the reactor was scrammed. 

Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in 

safe shutdown conditions. Capabilities of the RELAP5 code were challenged in this transient. The 

calculated values of the parameters show good agreement with the measured data. 

A parametric study was performed In order to evaluate the dependence of the steam generator level on the 

injected auxiliary feedwater flow. It indicated that the turbine driven auxiliary feedwater pump could possibly 

inject at a higher flowrate than its nominal value. 

The work was performed by the Department of Nuclear Engineering, Chalmers University of Technology in 

the framework of the Ringhals-3 power uprate project, supported by the Swedish Radiation Safety Authority 

(SSM). The ultimate goal of this project is to perform independent safety analyses of some limiting 

transients associated to the power uprate. The work carried out so far was targeted towards the 

development of state-of-the-art modelling capabilities for the Ringhals-3 unit. 

The present validational study is a Swedish contribution to the international Code Assessment and 

Maintenance Program (CAMP). 
iv 
v 

CONTENTS 

Page

ABSTRACT .................................................................................................................................... III

CONTENTS .................................................................................................................................... V

FIGURES ..................................................................................................................................... VII

TABLES ...................................................................................................................................... VIII

FOREWORD .................................................................................................................................. IX

ACKNOWLEDGMENT ..................................................................................................................... X

ABBREVIATIONS ........................................................................................................................... XI

1. INTRODUCTION .................................................................................................................... 1-1

2. DESCRIPTION OF THE TRANSIENT ...................................................................................... 2-1

3. FEATURES OF THE RELAP5 MODEL .................................................................................... 3-1

3.1 Status of the Model Development .................................................................................. 3-1

3.2 Description of the Nodalization ....................................................................................... 3-1

3.2.1 The downcomer .................................................................................................. 3-4

3.2.2 The lower plenum ............................................................................................... 3-4

3.2.3 The core inlet ...................................................................................................... 3-4

3.2.4 The core ............................................................................................................. 3-4

3.2.5 The bypass channels ......................................................................................... 3-6

3.2.6 The upper plenum .............................................................................................. 3-6

3.2.7 The hot and cold legs ......................................................................................... 3-6

3.2.8 The pressurizer .................................................................................................. 3-6

3.2.9 The SG primary side........................................................................................... 3-6

3.2.10 The main circulation pumps ............................................................................... 3-7

3.2.11 The safety systems ............................................................................................ 3-7

3.2.12 The feedwater system ........................................................................................ 3-7

3.2.13 The Steam Generators ....................................................................................... 3-7

3.2.14 The Steam Line .................................................................................................. 3-7

3.2.15 Turbines and steam dump systems.................................................................... 3-9

4. RESULTS OF THE BASE CASE CALCULATIONS ................................................................... 4-1

4.1 Reaching steady-state .................................................................................................... 4-1

4.2 Transient results ............................................................................................................. 4-7

5. PARAMETRIC STUDY ON THE AUXILIARY FEEDWATER FLOW ........................................... 5-1
vi 

6. RUN STATISTICS ................................................................................................................. 6-1

6.1 Remarks on the SNAP Graphical Tool ........................................................................... 6-3

7. CONCLUSIONS ..................................................................................................................... 7-1

8. REFERENCES ....................................................................................................................... 8-1
vii 

Figures 

Page

1. The Ringhals Nuclear Power Plant with 1 BWR and 3 PWR units ........................................ 1-1

2. Nuclear facilities in Northern Europe ..................................................................................... 1-2

3. Narrow range levels in the upper part of SG 2 ...................................................................... 2-1

4. Nodalization of the reactor pressure vessel internals ............................................................ 3-2

5. Nodalization of the primary side ............................................................................................ 3-3

6. Numbering scheme of the fuel channels in radial sectors ..................................................... 3-4

7. Radial mesh points and material compositions in the fuel elements ..................................... 3-5

8. Relative axial power distribution ............................................................................................ 3-5

9. Nodalization of the secondary side ....................................................................................... 3-8

10. Steady-state search of primary loop flowrate ....................................................................... 4-2

11. Steady-state search of feedwater mass flowrate ................................................................. 4-2

12. Steady-state search of steam mass flowrate ....................................................................... 4-3

13. Steady-state search of narrow range SG level .................................................................... 4-3

14. Steady-state search of pressures in the steam line ............................................................. 4-4

15. Steady-state PRZ pressure provided by a boundary volume ............................................... 4-4

16. Steady-state search of coldleg temperatures ....................................................................... 4-5

17. Steady-state search of hotleg temperatures ........................................................................ 4-5

18. Feedwater mass flowrate to the malfunctioning SG-2 ......................................................... 4-7

19. Narrow range level in the malfunctioning SG-2 .................................................................... 4-8

20. Wide range level in the malfunctioning SG-2 ....................................................................... 4-8

21. The measured neutronic power and the relative heating power .......................................... 4-9

22. Pressure in the intact SG-1 ................................................................................................ 4-10

23. Pressure in the malfunctioning SG-2 ................................................................................. 4-10

24. Steam mass flowrate in steamline 1 .................................................................................. 4-11

25. Steam mass flowrate in steamline 2 .................................................................................. 4-12

26. Narrow range level in the intact SG-1 ................................................................................ 4-13

27. Narrow range level in the intact SG-3 ................................................................................ 4-13

28. Auxiliary feedwater flow to the intact SG-1 (given as BC) .................................................. 4-14

29. Auxiliary feedwater flow to the intact SG-3 (nominal value is given as BC) ....................... 4-14

30. Coldleg temperature in the intact loop 1 ............................................................................ 4-16

31. Coldleg temperature in the malfunctioning loop 2 .............................................................. 4-16

32. Hotleg temperature in the intact loop 1 .............................................................................. 4-17

33. Hotleg temperature in the malfunctioning loop 2 ................................................................ 4-17

34. Level in the pressurizer ...................................................................................................... 4-18

35. Pressure in the pressurizer ................................................................................................ 4-18

36. Arrangement of the normal and auxiliary feedwater system ................................................ 5-1

37. AFW mass flowrate as varied parameter (boundary condition) ........................................... 5-2

38. SG-3 level with various AFW injections ............................................................................... 5-2

39. CPU time .............................................................................................................................. 6-2

40. Actual and Courant time steps ............................................................................................. 6-2

41. Mass error ............................................................................................................................ 6-3

42. Animation mask with fluid conditions and temperatures ...................................................... 6-4

43. Animation mask with void fractions ...................................................................................... 6-5
viii 

Tables

Page 

1. Sequence of events during the transient ............................................................................... 2-2

2. Comparison of calculated and measured steady-state parameters ...................................... 4-6

3. Summary of the components in the model ............................................................................ 6-1
ix 

FOREWORD 

The present code assessment report was prepared with the intention to provide feedback to the developers 

and the user community of the NRC computer codes in order to demonstrate the capabilities of 

RELAP5/Mod3.3 Patch 3. The authors made intensive efforts to build an up-to-date model of the 

Ringhals-3 unit for analyses of transients using best estimate methods. As the results reveal in the current 

document, the code has successfully been validated against an operational transient that occurred in the 

relevant NPP. 

I am convinced that the modeling strategies and main conclusions in this study will contribute to better 

understanding of a number of key thermal-hydraulic phenomena and will broaden the knowledge-base of 

safety authorities, as well as that of the code users worldwide. 

 Prof. Imre Pázsit 

 Department of Nuclear Engineering 

 Chalmers University of Technology 

 Gothenburg, Sweden 
x 

ACKNOWLEDGMENT 

The following people are acknowledged for the help/feedback/comments they provided during the 

completion of the calculations in the relevant project: Magnus Holmgren and Kjell Ringdahl from Ringhals 

AB; Anders Sjöberg from Studsvik Nuclear, Oddbjörn Sandervåg and Ninos Garis at SSM. 
xi 

ABBREVIATIONS 

AFW Auxiliary Feedwater 

ANS American Nuclear Society 

BC Boundary Condition 

BWR Boiling Water Reactor 

CAMP Code Assessment and Maintenance Program 

CL Cold Leg 

DC Downcomer 

ECC Emergency Core Cooling 

FA Fuel Assembly 

FW Feedwater 

HL Hot Leg 

LONF Loss of Normal Feedwater 

MSIV Main Steam Isolation Valve 

NPP Nuclear Power Plant 

NR Narrow Range 

NSSS Nuclear Steam Supply System 

PLS Precautions, Limitations and Setpoints 

PWR Pressurized Water Reactor 

PRZ Pressurizer 

RELAP Reactor Excursion and Leakage Analysis Program 

RHR Residual Heat Removal 

RPV Reactor Pressure Vessel 

SG Steam Generator 

SKI Swedish Nuclear Power Inspectorate 

SL Steam Line 

SNAP Symbolic Nuclear Analysis Package 

SSM Swedish Radiation Safety Authority (formerly known as SKI) 

WR Wide Range 
xii 
1-1 

1. INTRODUCTION

Many power utilities worldwide are applying for power uprates, i.e. for increasing the power output 

of their reactors. The Swedish Radiation Safety Authority (SSM, formerly known as the Swedish 

Nuclear Safety Inspectorate, SKI) received applications for new power uprates in Sweden, and 

among others, an application related to the Ringhals-3 Pressurized Water Reactor (Figure 1) in 

March 2004 [1]. This unit is a 3-loop Westinghouse type reactor, situated at the western coast of 

Sweden, approx. 60 km to the south of Gothenburg (Figure 2) [2]. 

Figure 1 The Ringhals Nuclear Power Plant with 1 BWR and 3 PWR units 

Increase of the thermal power in Ringhals-3 was planned to be carried out in two steps. During the 

first step, the thermal power will be increased from 2783 MWth (e.g. the original NSSS power level, 

100%) to 3000 MWth (108%). For the second step of the power uprate, the thermal power will be 

increased from 3000 MWth (108%) to 3160MWth (113.5%). 

It is essential to identify the main consequences resulting from the increased power level of the 

reactor, and their impact on the safety of the plant. SSM has to decide whether such power uprates 

still fulfill the requirements for a safe operation of the Ringhals-3 PWR. Performing an independent 

safety analysis of the Ringhals-3 power uprate was thus given by SSM as a research project to the 

Department of Nuclear Engineering, Chalmers University of Technology. 

Before stepping forward with the safety evaluation at an uprated power, a thermal-hydraulic model 

had to be created, updated and verified for the plant. With this objective in mind, a real operational 

transient, a Loss of Normal Feedwater case was chosen as a basis for validation of the model. By 

demonstration of the achievements and the shortcomings of the simulation, the authors intended to 

contribute to broadening of the knowledge base of the safety authority, as well as of the code user 

community. 
1-2 

Figure 2 Nuclear facilities in Northern Europe 
2-1 

2. DESCRIPTION OF THE TRANSIENT 

The feedwater control valves of the Ringhals-3 Unit were retrofitted during the outage in 2005 and, 

among others they were equipped with new digital valve positioners. The feedwater valves were 

tested after the outage and everything worked properly. On August 16th 2005, however, an 

electrical filter in the position transducer malfunctioned and therefore, the control signal to the valve 

failed. In such a case the valve immediately closes since that is the safe position of the valve. As a 

result, the normal feedwater (FW) flow to the steam generator (SG) becomes blocked. 

Initially, the lack of feedwater of SG-2 resulted in an increased amount of feedwater to the other two 

SGs before the safety and control systems were actuated. The steam production of the SG-2 was 

temporarily higher due to practically zero water inflow, while the heating power remained the same 

from the primary side. This change was represented by an increased primary system temperature 

and consequently, a higher pressurizer (PRZ) level and pressure. 

The level in the malfunctioning SG started to drop rapidly. At 38 s after shutting off the FW valve it 

reached the setpoint of the reactor trip, which is 12.4 % of narrow range span (Figure 3). The 

SCRAM signal triggered the control rods mechanism and the reactor was shut down. The trip logic 

and the control system provided a standardized procedure for such a case, including the trip of the 

turbines, and a simultaneous actuation of the auxiliary feedwater system. 

0

20

40

60

80

100

-50 0 50 100 150 200 250

SG Level [%]

Time [s]

SG 1

SG 2

SG 3

Narrow Range Span (100 %)

Nominal Level

12.4 %

38 s

Figure 3 Narrow range levels in the upper part of SG 2 
2-2 

There are three pumps operating in the auxiliary FW system: two electrical pumps and one steamdriven pump. Each pump feeds one separate SG. The steam-driven pump delivers twice as much 

flow as the motor-driven ones, and it is connected to the steam generator three. 

Shutting down of the reactor was followed by a turbine trip. Consequently, the pressure increased in 

the SG secondary side. With much less steam outflow, the temperature will also rise in the primary 

side. Therefore, the turbine bypass system actuates the steam dumping directly to the condenser in 

order to bring the temperatures back to normal. 

The reactor SCRAM resulted in a radical decrease of the heat input to the primary side. The steam 

content of the SG secondary side started to drop. The auxiliary FW system provided the necessary 

amount of heat sink, which was adequate for removal of the residual heat and getting the SG levels 

back to normal. Finally, the plant was stabilized in a hot standby mode. All the systems behaved as 

postulated after the initiating event. The sequence of the main events is summarized in Table 1. 

Table 1 Sequence of events during the transient 

Time [s] Event 

0 FW control valve of SG-2 is closing due to malfunction 

38 SG-2 level reached 12.4 % of narrow range: setpoint for SCRAM 

38 Turbine control valve is closing 

38 Auxiliary FW is actuated 

39 Steam dump valve starts opening 

47 Loop average temperature drops below 295 C 

47 The intact SG-1 and SG-3 FW control valves start closing 

77 Steam dump valves are closed 

250 End of simulation 
3-1 

3. FEATURES OF THE RELAP5 MODEL

3.1 Status of the Model Development

The current model of Ringhals-3 was built on the basis of a legacy input created by Studsvik 

EcoSafe [3] for an earlier version of RELAP5 in 1994. Obviously, the RELAP5 code itself has gone 

through extensive changes during this period. Therefore, the model had to be thoroughly modified 

in order to comply with the syntax rules of the new code version, and to represent the current 

situation at Ringhals. But the most crucial goal was to provide a very detailed core description in 

accordance with a simultaneously developed neutron kinetic model for the PARCS code [4]. The 

code version used in this study is RELAP5/Mod3.3 Patch 3-hh, the latest version officially released 

under the CAMP user agreement at the time of preparation this document. 

The majority of syntactical changes were related to the new volume and junction control flags, and 

the new connection coding requirements. The description of cross-flow junctions is easier and more 

logical with the application of the new expanded connection codes. Since the Henry-Fauske critical 

discharge flow model became the default option for the valve components in Mod 3.3, the 

corresponding valve parameters were updated accordingly. 

The development of the model was carried out in successive steps. At first, the original structure of 

the cold leg, including the ECC mixer components were modified, following the explicit 

recommendations of the User’s Guidelines [5]. Further structural changes were necessary on the 

secondary side. In order to capture the pressure pulse propagation, the number of sub-volumes of 

the steam line was increased by a factor of 10. 

The largest modifications were related to the reactor pressure vessel internals. The initial model 

had a very coarse nodalization, including only one downcomer and two channels in the core. One 

“hot channel” represented the hottest fuel bundle, and one “average channel” represented all the 

rest of the active core. Therefore, the most significant refinement was done in the core region by 

modeling each of the 157 fuel assemblies individually, both for the hydrodynamics and for the heat 

structures. The main objective of the full radial description of the core is to study potential 

occurrence of asymmetric behavior during some certain transients. An attempt was made to apply 

as many axial nodes as available in the neutron kinetic calculation (altogether 24). However, this 

has exceeded the capacity of RELAP5. Consequently, the axial discretization of the active core was 

kept the same as in the old model, i.e. using 8 vertical nodes. 

Results of the modeling studies have been summarized in various conference papers and technical 

reports [6], [7], [8], [9], [10], [11], [12], [13]. 

3.2 Description of the Nodalization

The RELAP5 nodalization schemes of the reactor pressure vessel (RPV) internals and the primary 

side are depicted in Figure 4 and Figure 5, respectively. Description of the most important parts of 

the model is given in the followings. The referenced component numbers appear in parentheses. 
3-2 

Figure 4 Nodalization of the reactor pressure vessel internals 

75

5

11 12 13

14 15 16

17 18 19 20 21 22

23 24 25

29 30 31

76 77 78

71

72

171

176

... 271

...

276

371

...

376

54

55

56

51

52

53

91

92

93

64

65

66

84

85

86

191

...

196

291

...

296

391

...

396

81

82

83

97

98

99

61

62

63

94

95

96

Active Core

Fuel Assemblies

67 68 69

4

3

CL1

CL2

CL3

HL1

HL2

HL3

with 157

Bypass 1: BaffleBypass 2: Flow in Open Guide

Thimbles and Gaps Between FAs

Bypass 3: Flow in Core Periphery

(Reflector Zone)

Barrel Space

Downcomer

Coldleg Connections

Hotleg Connections

and 8 Axial Nodes
3-3 

Figure 5 Nodalization of the primary side 440

205 206 207

330

326 331

320 340

305

390 386 381 383 385 905

928

495

497

498

442

441

350 496

906

461

463

462

460

945

941

942

940

918

916

915

910

943

932

930

931

933

405

435

220 240

226

126

431

231

131

230

283

290

285

286

281

904

925

493 250

947

924

903

120 140

130

105

450

927

948

316 346

216

246

116 146

150 491

922

902

190 186 183 185 181

492

901

946

921

451

494

911

Loop 3

Loop 2

Loop 1

SG 3

SG 2

SG 1

Pressurizer

Accu 3

Accu 2

Accu 1

Boron Tank

Charging Line

RHR 2

RHR 1

Normal Letdown

Safety Valve

ECCMix

ECCMix

ECCMix

ECCMix

ECCMix

ECCMix

PORV

Pump

Pump

Pump

75

5

11 12 13

14 15 16

17 18 19 20 21 22

23 24 25

29 30 31

76 77 78

71

72

171

176

... 271

...

276

371

...

376

54

55

56

51

52

53

91

92

93

64

65

66

84

85

86

191

...

196

291

...

296

391

...

396

81

82

83

97

98

99

61

62

63

94

95

96

67 68 69

4

3

Bypass Channels

Active Core

Fuel Assemblies

with 157

and 8 Axial Nodes

380

280

180
3-4 

3.2.1 The downcomer 

It is essential that some specific phenomena, for instance an asymmetric behavior of the loops 

should be captured properly. For this reason, the downcomer and the core were split into 3 parallel 

channels in order to retain the 3-loop structure of the primary side even within the reactor pressure 

vessel. The coldlegs are connected to the top of the downcomer via branches 011, 012, and 013. 

The downcomer is modeled with 3 annulus components (014, 015, and 016) representing 1/3 of the 

total flow area, each divided into 6 axial volumes. 

3.2.2 The lower plenum 

The large volume at the bottom of the RPV is modeled with altogether 6 volumes, 3 out of them 

representing the down-flow region (017, 18, and 019) and the other 3 represent the upwards flow 

part (020, 021, and 022). 

3.2.3 The core inlet 

The bottom of the core region consists of the lower support plate, the bottom nozzle, and the core 

plate models (23...28). The core inlet and outlet needed special considerations due to a limitation in 

RELAP5, since a branch component may be connected to maximally 9 other volumes from its each 

face. Thus, arrangement of 157 junctions to the fuel assemblies was realized by application of 

altogether 18 branches (171...76, 271...276 and 371...376). 

3.2.4 The core 

From hydrodynamic point of view, the fuel assemblies are modeled with vertical pipe components, 

axially divided into 8 volumes. It is assumed that 52 channels belong to one radial “sector” of the 

core (Figure 6). One loop is associated with one radial sector, i.e. 1/3 of the core. The fuel 

assembly models are basically identical, except the channel at the centre of the core (364). 

1

1

2

2

3

3

4

4

5

5

6

6

7

7

8

8

9

9

10

10

11

11

12

12

13

13

14

14

15

15

301 302 304 308 309 310 364 303

311 312 313 314 315 317 318 319

321 322 323 324 325 327 328 329

332 333 334 335 336 337 338 339

341 342 343 344 345 347 348

349 351 352 353 354 355

356 357 358 359 360

361 362 363

201 202 203

204 208 209 210 211 212 213

214 215 217 218 219 221 222 223

224 225 227 228 229 232 233 234 235

236 237 238 239 241 242 243 244 245

247 248 249 251 252 253 254 255

256 257 258 259 260 261 262 263

101

102 103

104 108 109

111 112 113 114 115

110

117 118 119 121 122 123 124

125 127 128 129 132 133 134

135 136 137 138 139 141 142

143 144 145 147 148

149 151 152 153 154

155 156 157 158

159 160 161

162 163

LOOP 1

LOOP 2

LOOP 3

Figure 6 Numbering scheme of the fuel channels in radial sectors 
3-5 

A fuel assembly is modeled with a heated channel. The heat source of the fuel rod is provided by 

either a control variable (in a stand-alone RELAP5 calculation), or by the neutron kinetic code (in a 

coupled RELAP5/PARCS simulation). A fuel element has 9 radial mesh points. Locations and the 

material compositions are shown in Figure 7. The relative axial power distribution is approached by 

a curve shown in Figure 8. 

1 2 3 4 5

6 7

8 9

1.83200

2.59000

3.17300

3.66300

4.09575

4.17830

4.46405

4.74980

Pellet Gap

Cladding

CL Distances from center line in mm

Figure 7 Radial mesh points and material compositions in the fuel elements 

 0

 0.05

 0.1

 0.15

 0.2

 0 0.5 1 1.5 2 2.5 3 3.5

Relative Power

Axial Height [m]

Relative Axial Power

Figure 8 Relative axial power distribution 
3-6 

3.2.5 The bypass channels 

It is also important to distinguish between the coolant flowing through the fuel assemblies while 

being heated, and the remaining part of the main loop flow. Thus, the core has been extended with 

altogether 3 bypass channels per loop, representing 

• the unheated baffle-barrel space, 

• the open guide thimbles, and 

• the flow path at the core periphery

The latter one is associated with the reflector zone explicitly modeled in the neutron kinetic code. 

The total amount of bypass flow was found to be playing an important role in the over-all core heat 

transfer. Consequently, the hydrodynamic parameters were carefully set in order to reproduce the 

measured ratio between the main flow and the bypass flow. Locations of the grid spacers were 

taken into account at setting the flow resistances and loss coefficients in the fuel bundle. Compared 

with plant data, these efforts resulted in a realistic pressure loss distribution over the core. 

3.2.6 The upper plenum 

The upper, inactive part of the core is simulated with 18 additional branch components in a similar 

manner as of the core inlet. The upper plenum consists of the core exit volumes (191...196, 

291...296 and 391...396) branching into 3 volumes (76, 77 and 78). The hot legs are connected to 

volume 71. The top of the RPV is modeled with a large volume (75). 

3.2.7 The hot and cold legs 

Loop 1 and loop 3 are essentially similar, while loop 2 is somewhat different because of the 

pressurizer. The length of the hot leg between the RPV outlet and SG inlet nozzle is L = 5.438 m, 

with an inner diameter of D = 0.7366 m. The length of the cross-over leg between the SG outlet and 

the pump inlet is L = 6.715 m, the diameter is D = 0.7873 m. The corresponding data for the cold 

leg (between the pump outlet and RPV inlet nozzle) are L = 7.331 m and D = 0.6985 m, 

respectively. 

3.2.8 The pressurizer 

The PRZ is modeled with 12 internal volumes in a pipe component (435). The lower 3 volumes are 

associated with electric heaters. The PORV (462), a safety valve (460), and the injection line of the 

PRZ spray system (440) are connected to the top volume of the PRZ. The mass flowrate of the 

sprayed coolant is controlled by a time dependent junction (442). The surge line (405) is branching 

from the cold leg (205) and it is connected to the bottom of the PRZ. 

Due to similarity of the 3 loops, the numbering scheme of the components is analogous within loop 

1, 2, and 3. Thus, x denotes 1, 2, or 3 in the following description of the components, respectively. 

3.2.9 The SG primary side 

The primary fluid enters the SG at the inlet nozzle and plenum (x20), consisting of 5 internal 

volumes. The vertical bundle of the heat exchanger tubes are modeled with one characteristic pipe 

(x30). Both legs of the U-tubes are partitioned into 13 vertical sub-volumes. The top of the bundle is 

represented by one horizontal sub-volume. The outlet plenum (x40) is symmetrical to the inlet. 
3-7 

3.2.10 The main circulation pumps 

The MCPs are modeled with pump components. The homologous curves, torque and head data 

were obtained from plant documentation. The model was not modified in the new version of the 

input deck. 

3.2.11 The safety systems 

One hydro-accumulator component (901, 903, and 905) is connected to each loop of the primary 

system through an emergency core cooling mixer component (x86). These ECC mixers are 

operating in tandem with the other mixers (x85) that may inject water either from the boron tank 

(916) or from the residual heat removal system (930 and 940). 

The primary side model is extended with a time dependent junction (498) simulating the normal 

letdown, and a charging line (450) and the corresponding valve (451). 

3.2.12 The feedwater system 

The boundary conditions for the normal feedwater are set in a time dependent volume (581). 

Distribution of the FW takes place in a branch component (854). The normal FW line consists of a 

control valve (862, 872, and 882) and an isolation valve (864, 874, and 884), respectively in each 

loop. The auxiliary FW is taken from the boundary volumes (891, 893, and 895). The mass flowrate 

of the injected auxiliary FW is controlled by time dependent junctions (892, 894, and 896). 

3.2.13 The Steam Generators 

The nodalization of the secondary side is shown in Figure 9. Similarities of the loops are utilized 

also in the case of numbering scheme of the SGs. Therefore, the internal structural elements of SG1, 2, and 3 are denoted with component numbers 5xx, 6xx, and 7xx, respectively. 

The FW enters to the SG at a branch (x76). The downcomer is modeled with a pipe component 

(x80) split into 5 vertical volumes. The large boiler section (x05) has altogether 14 axial volumes 

with internal diameter of 3.105 m. 

The tube support grids are taken into account. The bundle interphase friction model is applied in 

this region. Separation of the steam takes place in a separator component (x40). The inlet, the 

outlet and the liquid fallback junctions are connected to appropriate branch components (x35, x50, 

and x72, respectively). A steam dryer (x55) is situated in the upper part of the SG. 

3.2.14 The Steam Line 

The steam line is split into 2 parts (x85 and x95), each partitioned to 10 volumes. The main steam 

isolation valve (x94) is connecting the 2 parts together. Relatively high resolution of meshing was 

chosen in order to capture propagation of pressure pulses along the steam lines. Altogether 6 

safety valves (x86...x91) and a relief valve (x92) are connected to the steam line, in parallel with 

each other. 
3-8 

Figure 9 Nodalization of the secondary side 

735

635

535

771

671

571

775

675

575

875 878

885 888

865 868

872

882

862

892

896

894

874

884

864

854

776

676

576

705

605

505

781

681

581

780

680

580

765 665 565

871

881 861

873

883 863

772

672

572

740

640

540

750

650

550

755

655

555

661

760

660

560

761

764

664

564

561

685

785

585

594

592

692

792

586

686

786

587

687

787

588

688

788

589

689

789

590

690

790

591

691

694

791

794

695

795

595

826 828

599

891 895 893

598

824

818

814

822

816

812

800 801

770

670

570

827

823

817

813 851

SG 1

SG 2

SG 3

Safety Valves

Safety Valves

Safety Valves

Steam Line

Steam Line

Steam Line

Dryer

Dryer

Dryer

Separator Separator Separator

Relief Valve

Relief Valve

Relief Valve

Turbine 1

Turbine 2

Turbine Dump 1

Turbine Dump 2

Feedwater

Downcomer Downcomer Downcomer

MSIV

MSIV

MSIV

Drain Drain Drain

Steam Dump (Atmosphere)

FeedAuxiliary

Auxiliary

Auxiliary

FeedFeedControl Valve

Feedwater

Feedwater

Feedwater

Control Valve

Control Valve

Isolation Valve

Isolation Valve

Isolation Valve
3-9 

3.2.15 Turbines and steam dump systems 

The steam lines are merging at a branch component (800). The Ringhals-3 unit has 2 turbines and 

4 turbine dumps. These components are modeled in a simplified way in the current RELAP5 model. 

Turbine 1 and 2, and their control valves are simulated with time dependent junctions (814 and 824) 

and motor valves (813 and 823), respectively. One turbine dump system model is a combination of 

2 real dump systems in the plant (time dependent volumes 818 and 828). 

The control and protection systems were designed carefully and the parameters strictly follow the 

values laid out in the Ringhals-3 PLS document [14]. Improvement in the modeling of the control 

system resulted in a more realistic simulation of the feedwater flow by controlling the speed of the 

feedwater pumps. The controls influencing the steam dump system was also updated on the basis 

of new specifications. 
3-10 
4-1 

4. RESULTS OF THE BASE CASE CALCULATIONS 

4.1 Reaching steady-state

The steady-state initialization took approximately 300 s, running the codes with “transient” option on 

card 100. Using this option instead of “steady-state” was necessary in order to avoid a premature 

termination of the calculation. (Earlier test-runs resulted in false indications of reaching steadystate, in spite of some key parameters were in fact varying). 

The following strategy was applied in for achieving steady-state. The calculated heat balance 

resulted in a particular primary loop flowrate. This mass flowrate was set as the prescribed quantity 

for the circulation pumps to be reached by variation of the pump speed. The level in the PRZ was 

achieved by letting the PRZ spray and heaters to be actuated according to their real control 

systems. On the secondary side, the necessary SG level was reached by controlling the feedwater 

flow. 

In order to demonstrate the steady-state search process by the computer code, the main 

characteristic parameters are plotted on the next few pages. 

Due to the pump speed control system, the primary side loop flowrate (Figure 10) stabilized very 

rapidly. Only relative value of this parameter was available in the measured database. Therefore, 

the corresponding numerical value (expressed in kg/s, instead of %) was determined from the heat 

balance. 

The level control system of the SGs regulated the amount of feedwater (Figure 11) and steam 

(Figure 12). As it can be seen in these figures, the stabilized values converged to approx. 492 kg/s. 

Due to loop asymmetry, the injected FW to SG-2 is approx. 1 kg/s higher. Minor discrepancy (less 

than 3 %) was found between the measured and calculated steady-state values of the FW. 

Nevertheless, this can be explained by the precursory event of the control valve malfunctioning 

before the transient initiation, which is not part of the steady-state simulation. 

Basically, identical amounts of FW and steam mass flow rates led to stable levels in the SGs. As 

Figure 13 indicates, the narrow range SG levels converged to approx. 69 %, and this value is in 

very good agreement with the measured data. 

Secondary side pressures are depicted in Figure 14. The stationary pressures approached the 

measured value with a good accuracy. Concerning SG-2, a minor asymmetry is naturally present 

also in this case. 

The primary system pressure is maintained by a boundary volume connected to the PRZ. It can be 

observed in Figure 15 that the necessary pressure is assured by this arrangement. Obviously, the 

boundary volume is eliminated at the transient initiation. 

Both the cold leg (Figure 16) and hot leg (Figure 17) temperatures stabilized exactly at their 

measured values. Thus, it can be concluded that the over-all calculated parameters became stable 

and reflected the measured plant data with a good degree of accuracy at the end of the current 

initialization process. Numerical values are summarized in Table 2. 
4-2 

 4650

 4660

 4670

 4680

 4690

 4700

 4710

 4720

 4730

 0 50 100 150 200 250 300

Mass Flowrate [kg/s]

Time [s]

R3 LONF: Mass Flowrates in the Primary

Loop 1

Loop 2

Loop 3

Figure 10 Steady-state search of primary loop flowrate 

 320

 340

 360

 380

 400

 420

 440

 460

 480

 500

 520

 540

 0 50 100 150 200 250 300

Mass Flowrate [kg/s]

Time [s]

R3 LONF: Feedwater Mass Flowrates

FW 1

FW 2

FW 3

Figure 11 Steady-state search of feedwater mass flowrate 
4-3 

 485

 490

 495

 500

 505

 510

 515

 0 50 100 150 200 250 300

Mass Flowrates [kg/s]

Time [s]

R3 LONF: Steam Mass Flowrates in the Steamlines

SL 1

SL 2

SL 3

Figure 12 Steady-state search of steam mass flowrate 

 64

 65

 66

 67

 68

 69

 70

 71

 72

 0 50 100 150 200 250 300

Level [%]

Time [s]

R3 LONF: Levels in SGs (Narrow Range)

SG 1

SG 2

SG 3

Figure 13 Steady-state search of narrow range SG level 
4-4 

 6.1e+006

 6.12e+006

 6.14e+006

 6.16e+006

 6.18e+006

 6.2e+006

 0 50 100 150 200 250 300

Pressure [Pa]

Time [s]

R3 LONF: Pressures in SLs

SL 1

SL 2

SL 3

Figure 14 Steady-state search of pressures in the steam line 

 1.55098e+007

 1.55099e+007

 1.551e+007

 1.55101e+007

 1.55102e+007

 1.55103e+007

 1.55104e+007

 1.55105e+007

 1.55106e+007

 1.55107e+007

 1.55108e+007

 0 50 100 150 200 250 300

Pressure [Pa]

Time [s]

R3 LONF: PRZ Pressure

RELAP5

Figure 15 Steady-state PRZ pressure provided by a boundary volume 
4-5 

 556

 556.2

 556.4

 556.6

 556.8

 557

 0 50 100 150 200 250 300

Temperature [K]

Time [s]

R3 LONF: Coldleg Temperatures

CL 1

CL 2

CL 3

Figure 16 Steady-state search of coldleg temperatures 

 592

 592.2

 592.4

 592.6

 592.8

 593

 0 50 100 150 200 250 300

Temperature [K]

Time [s]

R3 LONF: Hotleg Temperatures

HL 1

HL 2

HL 3

Figure 17 Steady-state search of hotleg temperatures 
4-6 

Table 2 Comparison of calculated and measured steady-state parameters 

Parameter Name in RELAP5 Unit 

Steadystate 

value 

RELAP5 

average 

Measured 

average 

Deviation 

[%] 

Mass flow in loop 1 mflowj-180010000 kg/s 4658.01

Mass flow in loop 2 mflowj-280010000 kg/s 4660.17 4658.30 4654.12* 0.0898

Mass flow in loop 3 mflowj-380010000 kg/s 4656.73

FW flowrate to SG 1 mflowj-868000000 kg/s 491.75

FW flowrate to SG 2 mflowj-878000000 kg/s 492.70 492.10 507.21 2.9796

FW flowrate to SG 3 mflowj-888000000 kg/s 491.85

Steam flowrate in SL 1 mflowj-594000000 kg/s 491.70

Steam flowrate in SL 2 mflowj-694000000 kg/s 492.64 492.05 507.88 3.1171

Steam flowrate in SL 3 mflowj-794000000 kg/s 491.82

Level in PRZ cntrlvar-430 % 43.49 43.49 43.24 0.5692

NR level in SG 1 cntrlvar-521 % 68.98

NR level in SG 2 cntrlvar-621 % 68.98 68.98 68.96 0.0268

NR level in SG 3 cntrlvar-721 % 68.98

Normalized power cntrlvar-002 - 1.00 1.00 1.00 0.00

Pressure in PRZ p-435120000 Pa 15510718 15510718 15506794 0.0253

Pressure in SL 1 p-585050000 Pa 6160558

Pressure in SL 2 p-685050000 Pa 6153947 6158397 6165568 0.1163

Pressure in SL 3 p-785050000 Pa 6160688

Temperature in CL 1 tempf-181010000 K 556.45

Temperature in CL 2 tempf-281010000 K 556.39 556.43 556.45 0.0029

Temperature in CL 3 tempf-381010000 K 556.45

Temperature in HL 1 tempf-105030000 K 592.26

Temperature in HL 2 tempf-205030000 K 592.42 592.31 592.54 0.0382

Temperature in HL 3 tempf-305030000 K 592.26

_____________ 

* From heat balance calculation 
4-7 

4.2 Transient results

As it was mentioned in the Introduction, the relevant transient started with malfunctioning of the 

guiding electronics of the FW control valve in the injection line connected to SG-2. The amount of 

FW flow deviated from its stable value about a minute prior to the eventual shut down of the valve. 

This precursory event appears as an oscillation in Figure 18. Since this period was not part of the 

transient simulation, the FW mass flowrate was approximated with its average steady-state value. 

The transient began with an instant loss of FW, which was modeled with a prompt closure of valve 

in the code. 

 0

 100

 200

 300

 400

 500

 600

-50 0 50 100 150 200 250

Mass Flowrate [kg/s]

Time [s]

R3 LONF: Feedwater Mass Flowrates to SG-2

Measured

RELAP5

Figure 18 Feedwater mass flowrate to the malfunctioning SG-2 

Sinking of the level started immediately as it can be seen in Figure 19, reaching the so-called “LowLow Level” reactor trip condition at approx. 34 s. The narrow range measurement supplies higher 

resolution data, however, it obviously indicates zero value when the level passes below its lower 

differential pressure transducer. Hence, the wide range measured data are more suitable for 

comparison after the reactor SCRAM (Figure 20). 

The code was able to reproduce almost the entire period of level decrease. Dropping of the level 

stopped at approx. 35 % of wide range data and was followed by a slow increase. However, 

regaining of the level took place with a lower gradient than in the measurement. This was partly due 

to the underestimated amount of AFW injected to SG-2, as will be explained later on, and it is also a 

subject of a parametric study. 
4-8 

 0

 20

 40

 60

 80

 100

-50 0 50 100 150 200 250

Level [%]

Time [s]

R3 LONF: Levels in SG-2 (Narrow Range)

Measured

RELAP5

Figure 19 Narrow range level in the malfunctioning SG-2 

 30

 35

 40

 45

 50

 55

 60

 65

 70

 75

 80

 85

-50 0 50 100 150 200 250

Level [%]

Time [s]

R3 LONF: Levels in SG-2 (Wide Range)

Measured

RELAP5

Figure 20 Wide range level in the malfunctioning SG-2 
4-9 

Because of reaching the “Low-Low Level” condition, the reactor was tripped. Most importantly, the 

code was able to capture the timing of the SCRAM signal. Concerning the reactor power, the 

measurement comes from power range monitoring ex-core detectors calibrated against calorimetric 

measurements at relatively large power levels. Such a calibration does not hold for low power 

levels, as the ones monitored after a reactor SCRAM (decay heat), thus explaining why the best 

approach was found to be simulating the decay heat curve with a boundary condition according to 

the well-established ANS-5.1 standard. 

Figure 21 shows the values of the relative power. The measured data represent the average of 4 

neutron flux transducers located at the same azimuth around the core but at two different axial 

elevations. The RELAP5 curve indicates also the relative amount of residual heat in the core after 

the SCRAM, with respect to the steady-state value. 

 0

 0.2

 0.4

 0.6

 0.8

 1

 1.2

-50 0 50 100 150 200 250

Power [%]

Time [s]

R3 LONF: Relative Neutronic and Heating Power

Measured

RELAP5

Figure 21 The measured neutronic power and the relative heating power 

Variation of the pressures in the (intact) SG-1 is depicted in Figure 22. Matching of the measured 

and simulated data is excellent. A small discrepancy can be observed soon after the initiation of the 

transient. Increase of the pressure took place in the secondary side when the turbine throttle control 

valves started to close to maintain the nominal power. Since the full turbine system, including the 

unknown throttling valve characteristics, is modeled in a simplified way, the pressure data are 

slightly different during a short interval between 15 and 37 s. 

Simulation of the pressure has been achieved with a same degree of accuracy in the malfunctioning 

SG-2. The code managed to reproduce the sudden increase of the pressure during the closure of 

the turbines, followed by dumping of steam into the condenser (Figure 23). 
4-10 

 6e+006

 6.2e+006

 6.4e+006

 6.6e+006

 6.8e+006

 7e+006

 7.2e+006

 7.4e+006

-50 0 50 100 150 200 250

Pressure [Pa]

Time [s]

R3 LONF: Pressures in SG-1

Measured

RELAP5

Figure 22 Pressure in the intact SG-1 

 6e+006

 6.2e+006

 6.4e+006

 6.6e+006

 6.8e+006

 7e+006

 7.2e+006

 7.4e+006

-50 0 50 100 150 200 250

Pressure [Pa]

Time [s]

R3 LONF: Pressures in SG-2

Measured

RELAP5

Figure 23 Pressure in the malfunctioning SG-2 
4-11 

Response of the system to such a large disturbance can induce asymmetric behavior in the loops. 

One example for such differences is the amount of steam produced by steam generators. The mass 

flowrate in steamline 1 is shown in Figure 24. 

-100

 0

 100

 200

 300

 400

 500

 600

-50 0 50 100 150 200 250

Mass Flowrates [kg/s]

Time [s]

R3 LONF: Steam Mass Flowrates in Steamline-1

Measured

RELAP5

Figure 24 Steam mass flowrate in steamline 1 

The steam flow started to decrease from the intact SG-1 soon after the transient initiation, reaching 

a value of approx. 480 kg/s. Just around the time of the reactor SCRAM, the steam production was 

reduced almost instantaneously. This was interrupted by a small increase a few seconds later and 

followed by a continued dropping. The simulation could not reproduce this phenomenon, but 

instead it calculated a monotonous decrease, down to a minimal amount of steam. 

Opening of the steam dump valves resulted in a minor increase of the mass flowrate in the 

steamlines at around 50 s. This was also captured by the code, as well as the small amount of 

steady flow for the rest of the transient. 

Oscillatory nature of the measured steam flow is probably owing to the fact that the flow 

measurement system is not calibrated for such a small amount (5 – 7 %) of the nominal steam 

mass flowrate in the final phase of the transient. 
4-12 

-100

 0

 100

 200

 300

 400

 500

 600

-50 0 50 100 150 200 250

Mass Flowrates [kg/s]

Time [s]

R3 LONF: Steam Mass Flowrates in Steamline-2

Measured

RELAP5

Figure 25 Steam mass flowrate in steamline 2 

In the case of the malfunctioning SG-2, the amount of steam was initially increasing due to the total 

lack of FW injection. This phenomenon was present in the calculation but RELAP5 overestimated 

the flowrate, reaching the maximum of approx. 550 kg/s. Nevertheless, timing of the sudden closure 

of the control valve was perfectly matched. The intermittent dump valve opening resulted in a peak 

mass flowrate above 200 kg/s. The code calculated half as much steam dumped to the condenser 

during this short period. This is an indication for necessity of further improvement of the steam 

dump control system in the model. 

Concerning the collapsed levels in the other two intact steam generators, performance of the code 

is sufficiently good (Figure 26 and Figure 27). The narrow range measurements were available for 

the whole duration of the transient in this respect, since the levels were maintained above the lower 

taps of the differential pressure transducers. 

The dynamic response of secondary side is really good until approx. 100 s for both of the intact 

SGs. Lower speed of the level increase may be an indication of underestimated amount of auxiliary 

FW. This theory was supported when the available measured database and the plant 

documentation were compared. According to the latter one, the capacity of the turbine-driven AFW 

pump is approx. 48 kg/s. On the contrary, the measured AFW flow signal showed “saturation” when 

it approached 30 kg/s. In order to rectify this possible measurement flaw, the nominal value of the 

AFW mass flowrate (i.e. 48 kg/s) was used throughout the calculation. However, even this higher 

flowrate was found to be inadequately low for reproducing the correct SG levels. 
4-13 

 0

 10

 20

 30

 40

 50

 60

 70

 80

-50 0 50 100 150 200 250

Level [%]

Time [s]

R3 LONF: Levels in SG-1 (Narrow Range)

Measured

RELAP5

Figure 26 Narrow range level in the intact SG-1 

 0

 10

 20

 30

 40

 50

 60

 70

 80

-50 0 50 100 150 200 250

Level [%]

Time [s]

R3 LONF: Levels in SG-3 (Narrow Range)

Measured

RELAP5

Figure 27 Narrow range level in the intact SG-3 
4-14 

-5

 0

 5

 10

 15

 20

 25

 30

-50 0 50 100 150 200 250

Mass Flowrate [kg/s]

Time [s]

R3 LONF: Auxiliary Feedwater Mass Flowrates to SG-1

Measured

RELAP5

Figure 28 Auxiliary feedwater flow to the intact SG-1 (given as BC) 

-5

 0

 5

 10

 15

 20

 25

 30

 35

 40

 45

 50

-50 0 50 100 150 200 250

Mass Flowrate [kg/s]

Time [s]

R3 LONF: Auxiliary Feedwater Mass Flowrates to SG-3

Measured

RELAP5

Figure 29 Auxiliary feedwater flow to the intact SG-3 (nominal value is given as BC) 
4-15 

Regarding some primary loop parameters, such as temperatures of the coldlegs (Figure 30 and 

Figure 31), the shapes of the curves are rather similar to those of the secondary side pressures 

shown in Figure 22 and Figure 23. This strong correlation is quite obvious because the temperature 

of the fluid in the returning part of the primary loop is determined by the corresponding saturation 

pressure in the SG. 

From temperature point of view, behavior of the model is excellent. As Figure 30 and Figure 31 

show, temperatures have been very well matched in the hotlegs, at both the initial and final phases 

of the transient. Minor discrepancies exist during the previously mentioned turbine throttling. 

Sudden increase of the temperature, as well as timing of the peak, has been captured properly. 

In the case of the malfunctioning loop 2, the calculated initial coldleg temperature is identical with 

that of the intact loop 1. The real initial temperature was somewhat lower (approx. 0.6 K) due to the 

precursory event clarified before. After 50 s, RELAP5 slightly overestimated the temperatures in 

both cases but the magnitude of the discrepancy was less than 0.4 K, which was well within the 

measurement accuracy. 

Concerning the hotleg temperatures (Figure 32 and Figure 33), the transient can be characterized 

by essentially 3 intervals: a nearly stationary period until the SCRAM, a large temperature drop, and 

a slow decrease of the temperature at the final phase. Basically, the code was capable of predicting 

the values very well during these stages. 

Only a short delay of a few seconds duration can be observed in the timing of the sudden 

temperature drop. Two factors have influenced this phenomenon. The temperature measurement 

sensors are installed in metal pockets. Most likely, thermal inertia of the metallic structure played a 

role in the delay of the measured signal. In particular, magnitude of the instant cooling should be 

considered, which exceeded 30 K within a few seconds. On the other hand, the exact locations of 

the temperature measurements are just approximately known, which may result in some transition 

time, as well. 

The 3 characteristic stages of the transient can be well distinguished in the parameters related to 

the pressurizer. The PRZ level (Figure 34) and pressure (Figure 35) were increasing after the 

transient initiation until the SCRAM signal was triggered. At this point both of the parameters 

decreased due to the significantly reduced heating power. With evolution of the events, lower 

primary system temperature resulted in a decreased volume, and this phenomenon is reflected in 

the PRZ level. During the final phase, the PRZ pressure was basically maintained, or even very 

slightly re-pressurized. 

RELAP5 has managed to reproduce the parameters in PRZ very well. Magnitude of the variables, 

as well as the timing of the main occurrences has been captured accurately. 
4-16 

 6e+006

 6.2e+006

 6.4e+006

 6.6e+006

 6.8e+006

 7e+006

 7.2e+006

 7.4e+006

-50 0 50 100 150 200 250

Pressure [Pa]

Time [s]

R3 LONF: Pressures in SG-1

Measured

RELAP5

Figure 30 Coldleg temperature in the intact loop 1 

 6e+006

 6.2e+006

 6.4e+006

 6.6e+006

 6.8e+006

 7e+006

 7.2e+006

 7.4e+006

-50 0 50 100 150 200 250

Pressure [Pa]

Time [s]

R3 LONF: Pressures in SG-2

Measured

RELAP5

Figure 31 Coldleg temperature in the malfunctioning loop 2 
4-17 

 555

 560

 565

 570

 575

 580

 585

 590

 595

-50 0 50 100 150 200 250

Temperature [K]

Time [s]

R3 LONF: Hotleg Temperatures in Loop-1

Measured

RELAP5

Figure 32 Hotleg temperature in the intact loop 1 

 555

 560

 565

 570

 575

 580

 585

 590

 595

-50 0 50 100 150 200 250

Temperature [K]

Time [s]

R3 LONF: Hotleg Temperatures in Loop-2

Measured

RELAP5

Figure 33 Hotleg temperature in the malfunctioning loop 2 
4-18 

 15

 20

 25

 30

 35

 40

 45

 50

-50 0 50 100 150 200 250

Level [%]

Time [s]

R3 LONF: PRZ Level

Measured

RELAP5

Figure 34 Level in the pressurizer 

 1.38e+007

 1.4e+007

 1.42e+007

 1.44e+007

 1.46e+007

 1.48e+007

 1.5e+007

 1.52e+007

 1.54e+007

 1.56e+007

 1.58e+007

 1.6e+007

-50 0 50 100 150 200 250

Pressure [Pa]

Time [s]

R3 LONF: PRZ Pressure

Measured

RELAP5

Figure 35 Pressure in the pressurizer 
5-1 

5. PARAMETRIC STUDY ON THE AUXILIARY FEEDWATER FLOW 

Figure 36 shows the schematics of the feedwater injections to the steam generators. Normal FW 

became unavailable for SG-2 at the transient initiation. Soon after that, the normal FW was isolated 

from the other two intact SGs as well. The motor driven auxiliary feedwater system is connected to 

SG-1 and SG-2. 

SG-1 SG-2 SG-3

Normal FW Normal FW Normal FW

Motor Driven AFW Motor Driven AFW Turbine Driven AFW

Figure 36 Arrangement of the normal and auxiliary feedwater system 

The mass flowrate signals from these AFWs remained within their respective limits and provided 

reliable values. However, the turbine driven AFW signal became unreliable, falsely indicating that 

the maximal injection flowrate never exceeded approx. 30 kg/s. 

During the calculations, the largest discrepancy was found between the measured and calculated 

values of the collapsed levels in the SG-3. Since the injected AFW flow and the corresponding SG 

level are strongly interrelated, it become necessary to investigate the effects of various AFW 

flowrates by a parametric study. 

With this objective in mind, a series of calculations was accomplished over a range of different 

injection mass flowrates prescribed as boundary conditions. The most suitable component for this 

purpose was the time dependent junction no. 896 connected to SG-3. The applied user given 

values (Figure 37) were as follows: 

1) 30 kg/s Failed (“saturated”) measured signal 

2) 48 kg/s 100 % of the nominal value (base case) 

3) 60 kg/s 125 % of the nominal value 

4) 72 kg/s 150 % of the nominal value 

It can be seen in Figure 38 that the nominal value of the AFW resulted in an underestimation of the 

level. The measured quantity was well bounded by 60 kg/s and 72 kg/s, i.e. the best matching of 

the level is achieved with injection between 125 % and 150 % of the nominal AFW flowrate for a 

longer duration. However, even the constant value of 72 kg/s caused slightly slower level increase 

at the end of the transient. 
5-2 

 0

 10

 20

 30

 40

 50

 60

 70

 80

 0 50 100 150 200 250

Mass Flowrate [kg/s]

Time [s]

R3 LONF: AFW Injection to SG-3

Failed measured signal: ~30 kg/s

Nominal: 48 kg/s

60 kg/s

72 kg/s

Figure 37 AFW mass flowrate as varied parameter (boundary condition) 

 0

 10

 20

 30

 40

 50

 60

 70

 80

 0 50 100 150 200 250

Level [%]

Time [s]

R3 LONF: Narrow Range Levels in SG-3

Measured

Nominal: 48 kg/s

60 kg/s

72 kg/s

Figure 38 SG-3 level with various AFW injections 
6-1 

6. RUN STATISTICS 

The Ringhals-3 model was imported to the SNAP graphical interface and it reported the following 

summary about the components in the system model (Table 3): 

Table 3 Summary of the components in the model 

Number of hydraulic components: 462 

 Accumulators 3 

 Annulus 7 

 Branches 100 

 ECC mixers 6 

 Pipes 205 

 Pumps 3 

 Single junctions 30 

 Single volumes 34 

 Separators 3 

 Time dependent volumes 16 

 Valves 45 

 

Control systems 627 

 Trips 144 

 Control blocks 280 

 Signal variables 185 

 General tables 18 

 

Heat Structures 200 

Materials 10 

Connections 1790 

 Control connections 751 

 Heat connections 365 

 Hydraulic connections 674 

Running of the 250 s transient consumed 559.529 s of CPU time on a PC with a dual core Intel 

processor with 3.0 GHz speed and 3 GB RAM under a 64-bit version of MS Windows Vista. 

Time-step data: Mass error data:

 Minimum: 1.769404E-02 s Mass error: 3.287760 kg 

 Maximum: 2.500000E-02 s Total mass: 493441.0 kg 

 Average: 2.499750E-02 s Mass error / Total 6.662921E-06 

 Requested: 2.500000E-02 s 

 Attempted: 22121 

 Successful: 22121 
6-2 

 0

 100

 200

 300

 400

 500

 600

 0 50 100 150 200 250

Time [s]

Time [s]

R3 LONF: CPU Time

CPU Time

Figure 39 CPU time 

 0

 0.1

 0.2

 0.3

 0.4

 0.5

 0.6

 0 50 100 150 200 250

Time Step [s]

Time [s]

R3 LONF: Time Steps Data

Actual DT

Courant DT

Figure 40 Actual and Courant time steps 
6-3 

Figure 39 indicates a smooth running transient, showing a nearly linear consumption of the CPU 

time. The Courant limit time-step (Figure 40) was just slightly higher than the actual time-step 

basically over the whole duration, except short periods between 75 – 85 s. Essentially, RELAP5 

was able to run with the requested time-step, i.e. 25 ms until the end. 

The calculated mass error (Figure 41) was less than 3.3 kg. This is practically negligible compared 

to the total system mass, because their ratio is in the order of magnitude of 10-6. 

There was not any sort of unphysical behavior or code failure experienced during accomplishment 

of the simulations. This was valid both for the command-line based execution of the code, and for 

using the graphical interface. 

-0.5

 0

 0.5

 1

 1.5

 2

 2.5

 3

 3.5

 0 50 100 150 200 250

Mass [kg]

Time [s]

R3 LONF: Mass Error

Mass Error

Figure 41 Mass error 

6.1 Remarks on the SNAP Graphical Tool

For utilization of the Symbolic Nuclear Analysis Package, the SNAP tool, simplified animation 

masks were created. SNAP was found to be very useful for graphical representation of the model 

parameters. It helped to visualize, for instance, the fluid conditions (Figure 42), liquid temperatures, 

and void distribution (Figure 43). Even if the components are not proportionally scaled, a SNAP 

animation is extremely helpful for the analyst in better understanding of the entire process 

throughout a transient, such as the current Loss of Normal Feedwater scenario. 
6-4 

Figure 42 Animation mask with fluid conditions and temperatures 
6-5 

Figure 43 Animation mask with void fractions 
6-6 
7-1 

7. CONCLUSIONS

In this report, a code validation work carried out at the Department of Nuclear Engineering, 

Chalmers University of Technology in the framework of the Ringhals-3 power uprate project was 

presented. The first phase of the project was to develop an up-to-date thermal-hydraulic model of 

the Ringhals-3 PWR. 

With this objective in mind, a legacy RELAP5 input deck was extensively modified in order to 

update the input deck in accordance with the latest version of the code. One of the major 

improvements of the model was the complete re-nodalization of the active core. The new 

nodalization features a one-to-one radial mapping with the neutronic core model. Such a 

nodalization makes the model very versatile since all possible transients can be investigated, even 

in case of asymmetry between the different loops. The updated model was then successfully 

benchmarked against measured plant data corresponding to steady-state operation of the reactor. 

The main purpose of the current validation task during transient condition was also to evaluate the 

capabilities of the model to reproduce the system response to a large disturbance originating from 

the secondary side and to test the performance of the control system and of the trip logic. 

Comparisons between calculations and measured data demonstrated that the RELAP5 model is 

able to reproduce the main plant parameters with a high level of fidelity. A few possible areas for 

improvement were also identified, such as improving the auxiliary feedwater injection, and better 

modelling the control logic of the steam dump valves. 

The calculations have also revealed that the applied amount of auxiliary feedwater flow was 

sufficient for bringing the plant to safe shutdown conditions throughout the course of a Loss of 

Normal Feedwater transient. However, regaining of SG level took place at a slower pace with using 

the nominal AFW flowrate as boundary condition. Consequently, the actual AFW flowrate was 

supposedly higher than its nominal value specified in the plant documentation. Thus, the nominal

AFW flowrate appears to be a conservatively underestimated quantity because, in fact, the turbinedriven AFW pump is capable of delivering more feedwater to SG-3. 
7-2 
8-1 

8. REFERENCES 

1 Image Bank of Vattenfall AB at http://www.vattenfall.com/

2 Maps at International Nuclear Safety Center: http://www.insc.anl.gov/pwrmaps/

3 Eriksson, J., “RELAP/MOD3 Base Model for the Ringhals 3/4 Plants”. Technical Report, 

Studsvik EcoSafe, STUDSVIK/ES-94/14, 1994. 

4 Stålek, M., Demazière, C., “Development and Validation of a Cross-Section Interface for 

PARCS”, Annals of Nuclear Energy, 35 (2008) 2397-2409. 

5 Information System Laboratories Inc., “RELAP5/MO3.3 Code Manual, Volume V: User’s 

Guidelines”, (2006), Rockville, Maryland, Idaho Falls, Idaho Falls. 

6 Bánáti, J., Demazière, C., Stålek, M., 2006, “Development of a Coupled PARCS/RELAP5 

Model of the Ringhals-3 PWR,” Proc. of the Int. Topical Meeting on Advances in Nuclear 

Analysis and Simulation (PHYSOR2006), Vancouver, British Columbia, Canada, 

September 10-14, 2006. 

7 Demazière, C., 2006, “Development of a Cross-Section Interface for PARCS,” Proc. of 

the Int. Topical Meeting on Advances in Nuclear Analysis and Simulation

(PHYSOR2006), Vancouver, British Columbia, Canada, September 10-14, 2006. 

8 Demazière, C., Bánáti, J., 2006, “Application of the Coupled PARCS/RELAP5 codes for 

the Analysis of Transients in Connection with the Ringhals-3 Power Uprate,” SKI report. 

9 Bánáti, J., Demazière, C., Stålek, M., 2007, “Development and Validation of a Coupled 

PARCS/RELAP5 Model of the Ringhals-3 PWR,” Proc. of the 15th International 

Conference on Nuclear Engineering (ICONE-15), Nagoya, Japan, April 22 - 26, 2007. 

10 Bánáti, J., Stålek, M., Demazière, C. and Holmgren, M.: “Analysis of a Loss of Feedwater 

Case at the Ringhals-3 NPP Using RELAP5/PARCS Coupled Codes”, 16th International 

Conference on Nuclear Engineering (ICONE-16), Orlando, FL, USA, May 11-15, 2008. 

Paper No.: ICONE16-48116. 

11 Bánáti, J., Stålek, M., Demazière, C.: “Main Steam Line Break Calculations Using a 

Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor”, 16th 

International Conference on Nuclear Engineering (ICONE-16), Orlando, FL, USA, May 

11-15, 2008. Paper No.: ICONE16-48702. 

12 Bánáti, J., Stålek, M., Demazière, C.: “Validational Exercises with RELAP5/PARCS 

Coupled Codes Using Ringhals-3 Plant Data”, International Conference on Reactor 

Physics, Nuclear Power: A Sustainable Resource (PHYSOR-08), Interlaken, Switzerland, 

Sept. 14-19, 2008. 

13 Stålek, M., Bánáti, J. and Demazière, C.: “Coupled RELAP5/PARCS Main Steam Line 

Break Calculations Before and After a Power Uprate of a PWR”, International 

Conference on Reactor Physics, Nuclear Power: A Sustainable Resource (PHYSOR-08), 

Interlaken, Switzerland, Sept. 14-19, 2008. 

14 Ringhals AB: “Precautions, Limitations and Setpoints for Nuclear Steam Supply and 

Balance of Plant Systems”, Document ID: 1611225, Version: 18.0, 2004. 
8-2