NUREG/IA-0240

 International 

Agreement Report

Sensitivity Analyses of a 

Hypothetical 6 Inch Break,

LOCA in Ascό NPP using 

RELAP/MOD3.2

Prepared by:

R. Pericas, L. Batet, and F. Reventόs

Department of Physics and Nuclear Engineering

Technical University of Catalonia

ETSEIB, Av. Diagonal 647, Pav. C

08028 Barcelona

Spain

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

September 2010 

Prepared as part of

The Agreement on Research Participation and Technical Exchange 

Under the International Code Assessment and Maintenance Program (CAMP)

 

Published by

U.S. Nuclear Regulatory Commission
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 NUREG/IA-0240

Sensitivity Analyses of a hypothetical 6 inch

break, LOCA in Ascó NPP using

RELAP5/MOD3.2

Manuscript Completed: August 2010

Date Published: September 2010

Prepared by

R. Pericas, L. Batet, F. Reventós

Department of Physics and Nuclear Engineering

Technical University of Catalonia

ETSEIB, Av. Diagonal 647, Pav. C

08028 Barcelona

Spain

Prepared for:

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

iii

ABSTRACT

A postulated 6 inch break Loss-of-Coolant Accident (LOCA) is analyzed for the Ascó I plant (a 3 

loop PWR Westinghouse design) using RELAP5/Mod3.2. Different scenarios are calculated, 

including the total loss of the High Pressure Injection System (HPIS). In this case, the maximum 

cladding temperature rises above the steady state value at full power but, before the end of the 

transient, temperatures return to normal values by the effect of the accumulators injection.

Passive heat structures in the reactor pressure vessel have been incorporated to the model (i.e. 

the pressure vessel walls and some internals). Calculations with the model including passive 

heat structures lead to more severe scenarios (i.e. higher cladding temperatures) but still 

accumulators action is able to return the temperatures to normal values.

Finally, sensitivity to the trip time of the reactor coolant pumps is analyzed. Results show that 

lower cladding temperatures result when the pump trip is delayed. 
iv
v

FOREWORD

This report represents one of the assessment/application calculations submitted in fulfillment of 

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

Seguridad Nuclear (CSN) and the US Nuclear Regulatory Commission (USNRC) in the form of 

Spanish contribution to the Code Assessment and Management Program (CAMP) of the 

USNRC, whose main purpose is the validation of TRACE code.

The CSN and UNESA (the association of the Spanish utilities), together with some relevant 

universities, have set up a coordinated framework (CAMP-Spain), whose main objectives are 

the fulfillment of the formal CAMP requirements and the improvement of the quality of the 

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

centers and the engineering companies

This report is one of the Spanish utilities contributions to the above mentioned CAMP-Spain 

program and has been reviewed by the AP-28 Project Coordination Committee for the 

submission to the CSN.

UNESA

December 2009
vi
vii

CONTENTS

Page

ABSTRACT.................................................................................................................................. iii

FOREWORD................................................................................................................................. v

EXECUTIVE SUMMARY ............................................................................................................. ix

ABBREVIATIONS........................................................................................................................ xi

ACKNOWLEDGEMENTS...........................................................................................................xiii

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

2 PLANT DESCRIPTION .....................................................................................................2-1

2.1 Reactor pressure vessel.............................................................................................2-1

2.2 Core............................................................................................................................2-2

2.3 Pressurizer .................................................................................................................2-2

2.4 Steam Generators......................................................................................................2-2

2.5 Reactor coolant pumps ..............................................................................................2-2

2.6 Turbine system...........................................................................................................2-3

3 MODEL OF THE PLANT...................................................................................................3-1

4 SCENARIO .......................................................................................................................4-1

5 BASE CASE RESULTS ....................................................................................................5-1

6 ALTERNATIVE SCENARIOS ...........................................................................................6-1

7 SENSITIVITY TO THE PASSIVE HEAT STRUCTURES .................................................7-1

8 SENSITIVITY TO RCP STOP TIME .................................................................................8-1

9 CONCLUSIONS................................................................................................................9-1

10 REFERENCES................................................................................................................10-1

Figures

Page

Figure 1. Diagram of the Ascó model for RELAP5. ...................................................................3-2

Figure 2. RPV nodalization scheme...........................................................................................3-3

Figure 3. Steam Generators nodalization scheme.....................................................................3-4

Figure 4. General safety injections systems nodalization. .........................................................3-5

Figure 5. LOCA position scheme ...............................................................................................4-1

Figure 6. Base Case. Pressure..................................................................................................5-2

Figure 7. Base Case. Nuclear and Turbine Power. ...................................................................5-3

Figure 8. Base Case. Reactor Coolant Pumps Speed...............................................................5-3

Figure 9. Base Case. Steam Generators AFW Flow. ................................................................5-4

Figure 10. Base Case. Steam Dump Mass Flow. ......................................................................5-4
viii

Figure 11. Base Case. Core Mass Flow. ...................................................................................5-5

Figure 12. Base Case. Pressurizer Level. .................................................................................5-5

Figure 13. Base Case. Vessel Collapsed Level.........................................................................5-6

Figure 14. Base Case. Discharge and Injected Mass Flows. ....................................................5-6

Figure 15. Base Case. Integrated Discharge and Injected Mass Flows. ...................................5-7

Figure 16. Base Case. HRCT (axial nodes from 1 to 6). ...........................................................5-7

Figure 17. HPIS Injected Mass Flow Rate for the three scenarios. ...........................................6-2

Figure 18. Primary Circuit Pressure for the three scenarios. .....................................................6-3

Figure 19. Break Mass Flow Rate for the three scenarios.........................................................6-3

Figure 20. Integrated Break Mass Flow Rate for the three scenarios........................................6-4

Figure 21. Vessel Collapsed Level in the three scenarios. ........................................................6-4

Figure 22. Void Fraction in the volume above the core (for the three scenarios). .....................6-5

Figure 23. Alternative Scenario I. Discharge and Injected Mass Flows.....................................6-5

Figure 24. Alternative Scenario I. Integrated Discharge and Injected Mass Flows....................6-6

Figure 25. Alternative Scenario II. Discharge and Injected Mass Flows....................................6-6

Figure 26. Alternative Scenario II. Integrated Discharge and Injected Mass Flows...................6-7

Figure 27. Alternative Scenario I. HRCT, axial nodes 4 to 6. ....................................................6-7

Figure 28. Alternative Scenario II HRCT, axial nodes 3 to 6. ....................................................6-8

Figure 29. Sensitivity to passive HS. Primary Circuit Pressure. ................................................7-2

Figure 30. Alternative Scenario II. Heat tansfer from passive HS to coolant. ............................7-2

Figure 31. Sensitivity to the modeling of passive HS. Break Mass Flow Rate...........................7-3

Figure 32. Sensitivity to the modeling of passive HS. Accumulators Water Flow......................7-3

Figure 33. Sensitivity to the modeling of passive HS. Core levels.............................................7-4

Figure 34. Sensitivity to the modeling of passive HS. HRCT, axial node 5 ...............................7-4

Figure 35. Sensitivity to the modeling of passive HS. HRCT, axial node 6 ...............................7-5

Figure 36. RCP pumps sensitivity. Core water flow...................................................................8-2

Figure 37. RCP pumps sensitivity. Discharge water flow. .........................................................8-2

Figure 38. RCP pumps sensitivity. Accumulators water flow.....................................................8-3

Figure 39. RCP pumps sensitivity. Core levels..........................................................................8-3

Figure 40. RCP pumps sensitivity. Vapor fraction in the volume above the core. .....................8-4

Figure 41. RCP pumps sensitivity. HRCT, Axial Node 5. ..........................................................8-4

Figure 42. RCP pumps sensitivity. HRCT, Axial Node 6. ..........................................................8-5

Tables

Page

Table 1: Main features of Ascó NPP..........................................................................................2-1

Table 2: Model Nodalization main statistics...............................................................................3-3

Table 3: Base Case chronology.................................................................................................5-2

Table 4: Alternative Scenarios chronology. ...............................................................................6-2
ix

EXECUTIVE SUMMARY

The Department of Physics and Nuclear Engineering of the Technical University of Catalonia 

(UPC) holds a large background in the use of thermal-hydraulic codes for the Safety Analysis of 

Nuclear Power Plants (NPP). 

A postulated 6 inch break Loss-of-Coolant Accident (LOCA) is analyzed for the Ascó I plant (a 3 

loop PWR Westinghouse design). Different scenarios, including the total loss of the High 

Pressure Injection System (HPIS) are calculated.

A first set of calculations using the present RELAP5 model of the plant has been performed. In 

these calculations the effect of the HPIS actuation has been investigated. In addition to the 

Base Case (two out of two HPIS trains on duty), two alternative scenarios have been analyzed 

involving the failure of one and two HPIS trains. It is stated that in the Base Case the scenario 

does not represent any problem for the safety of the plant. In case of failure of one HPIS train, 

the cladding temperature of the upper part of the fuel elements may rise slightly but soon it is 

returned to the safety values by the action of the accumulators. In the case of malfunction of 

both HPIS trains, the cladding temperature rises above the steady state value but the injection 

from the accumulators returns the critical variables to normal values before the end of the 

transient.

Next step was a sensitivity analysis to assess the importance of including passive heat 

structures in the plant model. In order to do so, some vessel heat structures have been 

incorporated to the model (i.e. the pressure vessel walls and some internals). A new analysis 

was performed, comparing the worst case of the scenarios studied (complete loss of the HPIS 

trains) with and without passive heat structures. It is demonstrated that the inclusion of these 

passive heat structures leads to a more severe (i.e. higher cladding temperatures) but still 

accumulators action is able to return the temperatures to normal values before the end of the 

time span considered for the transient.

Finally, the sensitivity to the trip time of the reactor coolant pumps is analyzed. The worst case 

of the scenarios studied (complete loss of the HPIS trains) with passive heat structures is

recalculated with varying reactor coolant pumps stop times. It is found that the more the pump 

trip is delayed the temperature excursions are milder and, eventually, disappear. 
x
xi

ABBREVIATIONS

AFW Auxiliary Feed Water

ANAV Asociación Nuclear Ascó-Vandellòs

AS1 Alternative Scenario 1

AS2 Alternative Scenario 2

BC Base Case

EOP Emergency Operation Procedure

HS Heat Structures

HPIS High Pressure Injection System

HRCT Hot Rod Cladding Temperature

INL Idaho National Laboratory

LOCA Loss of Coolant Accident

MOL Middle-of-Life

NPP Nuclear Power Plant

PCT Peak Cladding Temperature

RCP Reactor Coolant Pumps

RELAP Reactor Excursion and Leak Analysis Program

SG Steam Generator

UPC Universitat Politècnica de Catalunya (Technical University of Catalonia)

UNESA Asociación Española de la Industria Eléctrica (Association of the Spanish 

Utilities)
xii
xiii

ACKNOWLEDGEMENTS

Authors are grateful to Association of the Spanish Utilities (UNESA) for funding the work 

presented in this report and to Asociación Nuclear Ascó-Vandellós (ANAV) for their continuous 

help and cooperation. Specially, authors recognize the valuable advices from Mr. Ismael Sol 

(ANAV).
1-1

1 INTRODUCTION

The Department of Physics and Nuclear Engineering of the Technical University of Catalonia 

(UPC) holds a large background in the use of thermal-hydraulic codes for the Safety Analysis of 

Nuclear Power Plants (NPP). More precisely, the Department‘s Thermal-Hydraulics Studies 

Group has been cooperating for 15 years with the operators of the Catalan nuclear plants, Ascó 

(two units) and Vandellós II [1].

In this document, a postulated medium size break Loss-of-Coolant Accident (LOCA) is analyzed 

for the Ascó I plant (a three loop PWR Westinghouse design). The selected transient consists in 

a 6 inch break in the cold leg, in different scenarios, including the total loss of the High Pressure 

Injection System (HPIS) flow (Sections five and six).

There are several previous studies of LOCA of this size or similar for the Ascó NPP. 

Nevertheless, the most comprehensive compendium is quite old (prior to the change of steam 

generators). 
2-1

2 PLANT DESCRIPTION

Ascó NPP has two units; each of them is a three-loop PWR of Westinghouse design.

The first unit is owned by ENDESA (100%). Second unit is owned by ENDESA (85%) and 

IBERDORLA (15%). The units are located close to Tarragona, in the north east of Spain, and 

they use the Ebro River as a final heat sink. The commercial operation of the plant started on 

December 1984. The actual nominal power of each unit is 2952.3 MWt and 1028 MWe. The 

reactor vessel is cold head type. The plant is equipped with the three Siemens (type SG 61 

W/D3) steam generators. The feed water is fed directly to the upper part of the downcomer via 

J-tubes. The circulation ratio on the secondary side of the steam generators is 3.65 at rated 

power. The auxiliary feed water system is pumped by one turbo pump and two motor pumps. In 

the plant there are, among others, control systems for the reactivity (rods and boron), primary 

pressure, pressurizer level, steam dump and steam generator level. The reactor protection 

system includes safety valves in the pressurizer and the steam generator.

2.1 Reactor pressure vessel

The vessel component of Ascó NPP corresponds to PWR three-loop Westinghouse nuclear 

power plant design. It has a cylindrical shape and the bottom is ended with a semi-spherical 

shape. The vessel contains the core, its support structure, control rods and also the thermal 

shield and other items directly associated with the core.

The vessel is designed and manufactured in accordance with the requirements of ASME Code 

concerning nuclear equipment. The material of construction of the vessel is low carbon steel 

alloy. The internal surfaces in contact with the cooling water, in order to minimize corrosion, are 

coated with a layer of austenitic stainless steel, deposited by welding.

Table 1: Main features of Ascó NPP.

Thermal reactor power (MWt) 2952.3

Electric power (MWe) 1028

Fuel UO2

Number of fuel elements 157

Loops 3

Reactor operation pressure (MPa) 15,51

Average coolant temperature(K):

Hot Zero Power

Hot Full Power

564,8

582,3

Steam generator Siemens

SG 61W/D3

Number of U-tubes in SG 5130

Total tube length (m) 98759

Inside tubes diameter (m) 0.0156

Tubes material INCONEL

Pumps type Westinghouse D 

100

Primary Circuit volume (m3) 106.19
2-2

Table 1: Main features of Ascó NPP. (continued)

2.2 Core

Ascó NPP core consists of 157 fuel assemblies with varying degrees of enrichment and burnup. Each fuel element contains 264 fuel rods, with a clad of Zircaloy-4 which contains in its 

internal UO2 pellets. The rod distribution matrix is 17x17. The number of control rods is 48. In 

every reload operation one third of the assemblies are changed.

2.3 Pressurizer

Pressurizer is standard Westinghouse design currently used in approximately 70 operating 

plants worldwide. The pressurizer volume is 1400 cubic feet, 39.64 m3. The heaters are 

vertically mounted, extending up through penetrations in the bottom head of the pressurizer 

shell. They are also individually seal-welded to the penetrations providing the system pressure 

boundary. The pressurizer heaters are one of the many components that have achieved very 

good overall performance in operating plants, they have a power about 377 kW, also there are 

extra heaters with a power about 1023 kW. Normal operating pressure is 15.51 MPa and normal 

operating temperature is 618 K. The maximum allowable pressure is 17.13 MPa, and the 

maximum allowable temperature is 633 K. 

2.4 Steam Generators

Original Westinghouse Steam Generators, were replaced by the present Siemens Steam 

Generators. The total primary side volume of the Steam Generators is 32.936 m3. The 

secondary side of the steam generators consists of a casing that acts as pressure barrier 

around the bundle of tubes and an upper zone that houses separators of steam moisture. The 

heat transferred by the reactor cooling water, through U-tubes raises the temperature of 

secondary until saturation, producing steam with high humidity content. Later this steam passes 

through a series of separators and dryers that remove virtually all water in the flow of steam 

toward the turbine. The water separated from the steam by separators, is mixed with main feed 

water.

2.5 Reactor coolant pumps

There are three reactor coolant pumps, one for each loop. Pumps model is W-1 1011-A1 (93-

DS). RCP is a vertical single-stage, centrifugal pump designed to move large volumes of reactor 

coolant water at elevated temperatures and pressures. Each pump consists of three general 

areas: the hydraulics, the seals and the motor. Pumps pressure is 8.6285 kg/cm2. 

Pressurizer volume (m3) 39.65

Pressurizer heaters power (kW) 1400
2-3

2.6 Turbine system

Turbine system is composed of two parts. In one side there is a one-body high pressure turbine, 

Westinghouse design. On the other side there is the low pressure turbine system, composed by 

two turbine modules, also Westinghouse design. They have a rotation speed of 1500 rpm.
3-1

3 MODEL OF THE PLANT

The RELAP5 code has been developed for best-estimate transient simulation of light water 

reactor coolant systems during postulated accidents. The code simulates the coupled behavior 

of the reactor coolant system and the core for loss-of-coolant accidents and operational 

transients such as anticipated transient without scram, loss of offsite power, loss of feed water, 

and loss of flow. A generic modeling approach is used that permits simulating a variety of 

thermal hydraulic systems. Control system and secondary system components are included to 

permit modeling of plant controls, turbines, condensers, and secondary feed water systems [2].

The Ascó NPP RELAP5 model is prepared to simulate both units of the plant. Only slight 

changes are needed, concerning mainly to the fuel load, to switch from one to another. When at 

full power, each plant produces, in the actual configuration, 2952.3 MW thermal (1028 MW 

electric). Although most of the main components of the plant are Westinghouse design, the 

present steam generators were designed by Siemens. 

The model of the plant includes hydrodynamic elements (primary, secondary, safety systems 

and auxiliary systems), heat structures, and control and protection systems. The model has 

been prepared for RELAP5/MOD3.2 and has been subjected to a thoroughly validation and 

qualification process, which includes the simulation of transients occurred in the plant itself 

[3],[4]. 

Figure one shows a general view of the hydrodynamic part of the model. The nodalization 

diagram for the reactor pressure vessel (RPV) is sketched in Figure 2 two, whereas 

Figure three reproduces the nodalization scheme used for the steam generators.
3-2

Figure 1. Diagram of the Ascó model for RELAP5.

As can be observed in figure one, High and Low Pressure Safety Injection Systems are 

connected to the primary system in the volumes numbered 270, 370 and 470. The modeling of 

the primary system safety injections is drawn in Figure 4, in which the two HIPS trains are 

indicated in red color. 

Regarding the core, the total number of fuel assemblies is 157. A fuel bundle consists OF a 

17x17 matrix of fuel pins, with 25 inactive positions for instrumentation and control rods. Active

core is 3,654 m high and it has a volume of water of 2,609 m3. In the model, the core is divided 

into six axial nodes, each one 0,609 m high.

Table two, summarizes the model’s degree of detail. During the preparation of the model, a 

great effort was devoted to the control and protection systems. Ascó model is able to reproduce 

the automatic response of the plant systems in practically all the circumstances and, besides, it 

incorporates some signals simulating operators’ actions.

HPI + LPI

HPI + LPI

Charge

CVCS

Discharge

CVCS

SG 1 SG 2 SG 3

Vessel

Spray

Relieve tank

Pressurizer

Accumulators

Loop 1 Loop 2 Loop 3

Turbine

Discharge to the condenser

MSR’s

Own consumption

220

225

320

325

420

425

281 381 481

449 450 460

465 470 475

349 350 360

365 370 375

249 250 260

265

251 252

270 275

530

535

538

 410

 

 310

 

 210

510 515

520

525

565

566

567

555

556

120

100

 104

130

810

431

610

231

645 745 845

230 240 330 340 430 440

Steam collector

Relieve valve: 653

Safety valves:

655, 657, 659, 661, 663

Relieve valve: 753

Safety valves:

755, 757, 759, 761, 763

Relieve valves: 853

Safety valves:

855, 857, 859, 861, 863

907

581

710

331

A.W. A.W. A.W.

Loop 3

Loop 2

Loop 1

ASCO N.P.P.

245

345

445

HPI + LPI
3-3

Figure 2. RPV nodalization scheme.

Table 2: Model Nodalization main statistics.

Component type Number of elements

Hydrodynamic volumes 549

Heat slabs 138

Heat structure nodes 559

Control variables 1454

Variable trips 219

Logical trips 431

Tables 241

160

155

109

110

112

100

104

120

130

140

108 158 150

410 310 210

 Hot Legs

Cold Legs
3-4

601

620

630

625

645

640

225

622

615

676

A.A.

235 236

from hot leg to cold leg

245

230 240

steam outlet

600

249

600 700 800

GV 1 GV 2 GV 3

225

230

235

240

245

601

610

615

620

625

630

640

650

325

330

335

340

345

701

710

715

720

725

730

740

750

425

430

435

440

445

801

810

815

820

825

830

840

850

primary secondary

231 331 431

622 722 822

231

610

220

605

606 SG drain

675

236 336 436

650

645 745 845

605 705 805

drain

606 706 806

Figure 3. Steam Generators nodalization scheme.

The original model incorporated heat structures on the following components:

- Steam Generators.

- Core.

- Pressurizer.

- Feed water heaters.

It is known that in this kind of scenarios (i.e. medium sized LOCA) the heat stored in passive 

structures has a noticeable influence in the results. In order to assess this premise, the 

calculation for the worst case of the analyzed (failure of both trains of the HPIS) is repeated in 

section 7, where passive heat structures are included in the model, and the results are 

compared. 
3-5

Figure 4. General safety injections systems nodalization.

Low pressure safety injection

Cold legs

loop 3

Accumulator loop1 

High pressure safety injection

Loop 2

Loop 1

010

011

018 019 020 021

022

028 029 030 031

032

033

034

035 036

213

214

038 039 040

041

048 049 050

051

060

061

070

071

042

043

052

053

055

215

216

315

316

415

416

Accumulator loop 2 Accumulator loop 3

313

314

413

414

281 381 481

480

479

478 476

380

379

378 376

280

279

278 276

270

370

470

218

318

418

ASCO NPP

carga

CVCS

306

305

303 304

301 302

477

377

277
4-1

4 SCENARIO

As it has been introduced above, the transient analyzed in this report is a postulated 6 inch

break Loss-Of Coolant Accident (LOCA) in Ascó-I NPP. It is assumed that the plant is operating 

at full power when the break occurs. The simulation has been carried out with the middle-of-life 

(MOL) conditions of the reload cycle 17.

The break is simulated as a sudden rupture at 50 seconds into the transient calculation. 

Immediately, reactor, turbine and the reactor coolant pumps trip.

The LOCA is postulated to take place in the cold leg, down-flow the reactor coolant pump (RCP) 

in loop one (the red colored square in Figure one indicates the location of the break). In figure 

five, more details of the simulation of the break are shown. The arrow indicates the place where 

the break is located. Volume 265 in figure five corresponds to the first pipe segment down-flow

the RCP and up-flow the Safety Injection point (all distances in millimeters). A new volume, 

numbered 291 and shown in figure five, has been added to the model to simulate the 

containment.

In order to run the transient, a previous steady state calculation has been necessary. It is worth 

to mention here that the only boundary conditions imposed on the steady state are, on the one 

hand, the characteristics of the fuel and the boron concentration (as said, Ascó I, Cycle 17, 

MOL), and in the other hand, the turbine admission pressure in the secondary. So, the model is 

allowed to run limited only by its own controls (which simulate the actual plant ones). In this 

case, the steady state calculation has been run for 1200 seconds.

For the transient, besides the break conditions (which is the only condition for the Base Case), 

some other restrictions have been imposed: for the Alternative Scenarios one and two one and 

two (of two) trains of the HPIS (depicted in red in figure four) have been considered nonavailable. 

Figure 5. LOCA position scheme

RCP

H&LPIS

Accumulator.

Vessel Inlet

813 3230 2831

LOCA 6’’

 291
5-1

5 BASE CASE RESULTS

In this section, the results for the Base Case simulation are presented (i.e. all the injection 

systems are in function). After the 6 inch break (at 50 s) takes place, a fast primary 

depressurization occurs and soon primary system decouples from the secondary one (figure 

six). Turbine and Reactor trip (figure seven), and also do RCPs (figure eight).

The Auxiliary Feed Water (AFW) enters almost immediately (5.1 seconds after the break) into 

function (figure nine) but there still exists an increase in the secondary pressure that forces the 

actuation of the Steam Dump (Figure 10). In figure nine it can be observed that AFW is far 

greater in SG2 than in the other two SG, this is due to the fact that AFW of this loop is driven by 

a turbopump while other loops use a motorpump.

The RCP coast down causes a decrease in the core mass flow to a value near to zero (Figure 

11). The oscillations around 450 seconds may be due to accumulator injection. See Figure 14

for accumulator injection.

The system depressurization is driven by the break flow. Figure 12 shows the level in the 

pressurizer.

Soon after the break, when the pressure reaches the set point, the HPIS starts injecting cold 

water in the three loops (Figure 14 and Figure 15). 

In Figure 13 it can be observed that in the Base Case no core uncovering occurs during the 

transient (collapsed level never gets lower than 3.7 m, which represents the Top of Active Fuel). 

Level recovery can be explained by comparing safety injection (accumulators plus HPIS) and 

discharge mass flows (Figure 14 and Figure 15).

Finally, the hot rod cladding temperature in the six core axial nodes is depicted in Figure 16. As 

can be seen in this plot, there is no rise of temperatures during the transient. This behavior will 

not be encountered in the alternative scenarios.
5-2

Table 3: Base Case chronology.

Event Time (s)

Break 50

RCP stop 50

Power stop 50

Turbine stop 50

Accumulator injection 330

HPIS injection 62

Hot channel peak cladding temperature, Node 5 none

Hot channel peak cladding temperature, Node 6 none

0 200 400 600 800 1000

0

20

40

60

80

100

120

140

160 Pressure (kg/cm

2rel.)

Time (s)

 Primary Circuit

 Secondary Circuit

Figure 6. Base Case. Pressure.
5-3

0 200 400 600 800 1000

0

20

40

60

80

100 Power (%)

Time (s)

 Core

 Turbine

Figure 7. Base Case. Nuclear and Turbine Power.

0 200 400 600 800 1000

0

20

40

60

80

100

120

140

160 Speed (rad/s)

Time (s)

 RCP1

 RCP2

 RCP3

Figure 8. Base Case. Reactor Coolant Pumps Speed.
5-4

0 200 400 600 800 1000

0

20

40

60

80

100

120

140

160

180

200 Flow (T/h)

Time (s)

 SG1

 SG2

 SG3

Figure 9. Base Case. Steam Generators AFW Flow.

0 200 400 600 800 1000

0

200

400

600

800

1000

1200

1400 Flow (T/h)

Time (s)

 Steam Dump

Figure 10. Base Case. Steam Dump Mass Flow.
5-5

0 200 400 600 800 1000

-5000

0

5000

10000

15000 Flow (T/h)

Time (s)

 Core

Figure 11. Base Case. Core Mass Flow.

0 200 400 600 800 1000

0

10

20

30

40

50

60 Level (%)

Time (s)

 Pressurizer

Figure 12. Base Case. Pressurizer Level.
5-6

0 200 400 600 800 1000

0

1

2

3

4

5

6

7

8

9

10

11

12 Level (m)

Time (s)

 Core

Top of active fuel (3,7m)

Figure 13. Base Case. Vessel Collapsed Level.

0 200 400 600 800 1000

0

200

400

600

800

1000

1200 Flow (kg/s)

Time (s)

 Discharge

 Acum+HPI

Figure 14. Base Case. Discharge and Injected Mass Flows.
5-7

0 200 400 600 800 1000

0

50000

100000

150000

200000

250000

300000

350000 Mass (kg)

Time (s)

 Discharge

 Accum + HPI

Figure 15. Base Case. Integrated Discharge and Injected Mass Flows.

0 200 400 600 800

440

460

480

500

520

540

560

580

600

620

640 Temperature (K)

Time (s)

 Node1

 Node2

 Node3

 Node4

 Node5

 Node6

Figure 16. Base Case. HRCT (axial nodes from 1 to 6).
6-1

6 ALTERNATIVE SCENARIOS

This section deals with the simulation of two alternative scenarios. In both cases the break is 

postulated to be the same that in Base Case. The difference among scenarios is that in the first 

alternative scenario (AS1) a malfunction of one HPIS train is postulated, whereas in the second 

alternative scenario (AS2), there is a total loss of the HPIS availability (Figure 17). In the Base 

Case (BC) HPIS is fully available.

In this section it is analyzed how partial or total unavailability of the HPIS modifies the 

progression of the scenario. Figures commented in this section use blue color for plots 

representing the worst of the scenarios (AS2), red color for the middle one (one train available) 

whereas black corresponds to the Base Case.

Figure 18 compares the primary circuit pressure for the three cases. In AS2 the pressure is the 

highest. Break flow (Figure 19 and Figure 20) is a slightly lower in the worst scenario because 

there is no HPIS injection on that scenario. 

Collapsed levels plotted in Figure 21 indicate that a core uncover can occur in both alternative 

scenarios, a fact that is corroborated by the void fraction value in the volume above the core 

(Figure 22). The plot corresponding to AS2 remains at its maximum longer than those of the 

other scenarios. When vapor fraction in the core nodes reaches values close to 1, cladding

temperatures also start rising. Figure 27 and Figure 28 show the hot rod cladding temperatures 

in the scenarios AS1 and AS2, respectively. In both cases the temperatures decrease again 

when the accumulators start injecting. In Figure 23 to Figure 26, the flow from the three 

accumulators is plotted for each of the alternative scenarios. 
6-2

Table 4: Alternative Scenarios chronology.

Event Alternative 

Scenario I

Alternative 

Scenario II

Discharge (s) 50 50

RCP stop (s) 50 50

Power stop (s) 50 50

Turbine stop (s) 50 50

Accumulators injection (s) 340 360

HPIS injection (s) 62 none

Hot channel peak cladding temperature, Node 5 (s) 360 380

Hot channel peak cladding temperature, Node 6 (s) 370 390

0 200 400 600 800 1000

0

50

100

150

200

250 Flow (T/h)

Time (s)

 BC

 AS1

 AS2

Figure 17. HPIS Injected Mass Flow Rate for the three scenarios.
6-3

0 200 400 600 800 1000

0

20

40

60

80

100

120

140

160 Pressure (kg/cm

2rel.)

Time (s)

 BC

AS1 AS2

Figure 18. Primary Circuit Pressure for the three scenarios.

0 200 400 600 800 1000

0

200

400

600

800

1000

1200

1400

1600

1800 Flow (kg/s)

Time (s)

 BC

AS1 AS2

Figure 19. Break Mass Flow Rate for the three scenarios.
6-4

0 200 400 600 800 1000

0

50000

100000

150000

200000

250000

300000

350000 Mass (kg)

Time (s)

 BC

 AS1

 AS2

Figure 20. Integrated Break Mass Flow Rate for the three scenarios.

0 200 400 600 800 1000

0

2

4

6

8

10

12 Level (m)

Time (s)

 BC

 AS1

 AS2

Top pf active fuel (3,7m)

Figure 21. Vessel Collapsed Level in the three scenarios.
6-5

0 200 400 600 800 1000

0,0

0,2

0,4

0,6

0,8

1,0 Vapor Fraction

Time (s)

 BC

AS1 AS2

Figure 22. Void Fraction in the volume above the core (for the three scenarios).

0 200 400 600 800 1000

0

500

1000

1500

2000 Flow (kg/s)

Time (s)

 Discharge

 Acum + HPIS

Figure 23. Alternative Scenario I. Discharge and Injected Mass Flows.
6-6

0 200 400 600 800 1000

0

50000

100000

150000

200000

250000

300000

350000

400000 Mass (kg)

Time (s)

 Acum + HPIS

 Discharge

Figure 24.. Alternative Scenario I. Integrated Discharge and Injected Mass Flows.

0 200 400 600 800 1000

0

500

1000

1500

2000 Flow (Kg/s)

Time (s)

 Discharge

 Accum + HPIS

Figure 25. Alternative Scenario II. Discharge and Injected Mass Flows.
6-7

0 200 400 600 800 1000

0

50000

100000

150000

200000

250000

300000

350000

400000 Mass (kg)

Time (s)

 Discharge

 Accum + HPIS

Figure 26. Alternative Scenario II. Integrated Discharge and Injected Mass Flows.

0 200 400 600 800 1000

400

450

500

550

600

650 Temperature (K)

Time (s)

 Node 4

 Node 5

 Node 6

Figure 27. Alternative Scenario I. HRCT, axial nodes 4 to 6.
6-8

0 200 400 600 800 1000

400

450

500

550

600

650 Temperature (K)

Time (s)

 Node 3

 Node 4

 Node 5

 Node 6

Figure 28. Alternative Scenario II HRCT, axial nodes 3 to 6.
7-1

7 SENSITIVITY TO THE PASSIVE HEAT STRUCTURES

As has been mentioned in Section three, the model used so far does not have passive heat 

structures (HS) in the vessel volumes. In this section a comparison is done in order to test how 

sensitive the calculation results are to the incorporation of these heat structures. So, passive 

heat structures have been added to the model in order to better capture the phenomena of the 

analyzed scenario

The new added heat structures include: the three vessel inlet nozzles (three), the downcomer 

walls (two), the lower and upper plenum walls (four), the core walls and the bottom and top 

vessel walls (eight). At the end, 17 new heat structures have been included in the model, with 

85 new heat structure nodes.

A new analysis has been performed, comparing the worst case of the scenarios studied 

(complete loss of the HPIS trains, AS2) with and without passive heat structures. To do so, a 

calculation has been run with the new model. 

From Figure 29 it can be observed that the case with passive HS yields a slightly higher

pressure inside the primary circuit. This difference is caused by the heat accumulated by the 

passive HS being released as the primary pressure decreases, providing an extra steam 

generation that reduces the depressurization rate (Figure 30). 

As can be observed in Figure 31, the break flow rate is quite similar in both cases. The time of 

accumulator injection, nevertheless, is different (Figure 32). Without the passive HS the 

accumulators start injecting slightly before (due to difference in primary pressure), but this 

difference is enough to cause a noticeable advancement in the level recovery (Figure 33).

Consequently, as can be stated from Figure 34 and Figure 35, the lack of the passive heat 

structures in the model leads to a quite non-conservative result.
7-2

0 200 400 600 800 1000

0

20

40

60

80

100

120

140

160 Pressure (kg/cm

2rel.)

Time (s)

AS2 AS2HS

Figure 29. Sensitivity to passive HS. Primary Circuit Pressure.

0 200 400 600 800 1000

0.00E+000

5.00E+007

1.00E+008

1.50E+008

2.00E+008

2.50E+008 Heat transference (W)

Time (s)

 Total heat transfer from passive HS

 Total nuclear power

Figure 30. Alternative Scenario II. Heat tansfer from passive HS to coolant.
7-3

0 200 400 600 800 1000

0

200

400

600

800

1000

1200

1400

1600

1800 Flow (kg/s)

Time (s)

AS2 AS2HS

Figure 31. Sensitivity to the modeling of passive HS. Break Mass Flow Rate.

0 200 400 600 800 1000

0

200

400

600

800

1000

1200 Flow (Kg/s))

Time (s)

AS2 AS2HS

Figure 32. Sensitivity to the modeling of passive HS. Accumulators Water Flow.
7-4

0 200 400 600 800 1000

0

2

4

6

8

10

12 Level (m)

Time (s)

AS2 AS2HS

Top of active fuel (3,7m)

Figure 33. Sensitivity to the modeling of passive HS. Core levels.

0 200 400 600 800 1000

400

450

500

550

600

650 Temperature (K)

Time (s)

AS2 AS2HS

Figure 34. Sensitivity to the modeling of passive HS. HRCT, axial node 5
7-5

0 200 400 600 800 1000

400

450

500

550

600

650 Temperature (K)

Time (s)

AS2 AS2HS

Figure 35. Sensitivity to the modeling of passive HS. HRCT, axial node 6
8-1

8 SENSITIVITY TO RCP STOP TIME

In this section a comparison is done in order to test how sensitive the results are to the trip time 

of the RCPs. A new calculation has been performed, comparing the worst case of the scenarios 

studied (complete loss of the HPIS trains) with passive heat structures when the reactor coolant 

pumps stop at different times. 

Several data are plotted in each graph. RCP no action means that pumps do not trip. 

RCPstop means that reactor coolant pumps trip at same time that LOCA occurs. 

RCP30 means that pumps are stopped 30 seconds after the LOCA. This selection has been 

chosen under “IOE-E-1 perdida del refrigerante del reactor o secundario” recommendations. 

IOE-E-1 is an EOP for Ascó NPP [5]. There, the instruction is to stop RCP pumps, when 

subcooling is lower than 0 oC. From the RCP no action scenario calculation, this condition 

happens at a time about 20 seconds into the transient. 10 more seconds have been added to 

account for operators actions delay. 

RCP70 means that pumps are stopped 70 seconds right after the LOCA. 

In Figure 36 it can be observed a comparison of the core mass flow for the different cases 

studied. Figure 37 shows how the break flow rate is quite similar in all cases. Accumulators start

to inject in different times (Figure 38), and this fact is enough to cause noticeable advancement 

in the level recovery (Figure 39 and Figure 40). Peak cladding temperature gets a quite nonconservative scenario for the RCP no action. The more the pump trip is delayed, the 

temperature excursions are milder and, eventually, disappear (Figure 41 and Figure 42). 
8-2

0 200 400 600 800 1000

-4000

-2000

0

2000

4000

6000

8000

10000

12000

14000 Flow (Kg/s)

Time (s)

 RCP no action

 RCPstop

 RCP30

 RCP70

Figure 36. RCP pumps sensitivity. Core water flow.

0 200 400 600 800 1000

0

200

400

600

800

1000

1200

1400 Flow (kg/s)

Time (s)

 RCP no action

 RCPstop

 RCP30

 RCP70

Figure 37. RCP pumps sensitivity. Discharge water flow.
8-3

0 200 400 600 800 1000

0

50

100

150

200

250

300

350

400 Flow (Kg/s)

Time (s)

 RCP no action

 RCPstop

 RCP30

 RCP70

Figure 38. RCP pumps sensitivity. Accumulators water flow.

0 200 400 600 800 1000

0

2

4

6

8

10

12 Level (m)

Time (s)

 RCP no action

 RCPstop

 RCP30

 RCP70

Top of active fuel (3,7m)

Figure 39. RCP pumps sensitivity. Core levels.
8-4

0 200 400 600 800 1000

0,0

0,1

0,2

0,3

0,4

0,5

0,6

0,7

0,8

0,9

1,0

1,1 Vapor Fraction

Time (s)

 RCP no action

 RCPstop

 RCP30

 RCP70

Figure 40. RCP pumps sensitivity. Vapor fraction in the volume above the core.

0 200 400 600 800 1000

450

500

550

600

650

700 Temperature (K)

Time (s)

 RCPno action

 RCPstop

 RCP30

 RCP70

Figure 41. RCP pumps sensitivity. HRCT, Axial Node 5.
8-5

0 200 400 600 800 1000

450

500

550

600

650

700 Temperature (K)

Time (s)

 RCP no action

 RCPstop

 RCP30

 RCP70

Figure 42. RCP pumps sensitivity. HRCT, Axial Node 6.
9-1

9 CONCLUSIONS

A 6 inch cold leg loss of coolant accident has been simulated for the Ascó 1 NPP at the

conditions of Middle of Life for the load cycle 17. The model of the plant, prepared for 

RELAP5/Mod3.2, has been thoroughly tested and validated (see [3] and [4]) and is considered 

fully qualified for this purpose.

From the calculations it is stated that, in this kind of accidents, if all the HPIS are on duty the 

scenario does not represent any problem for the safety of the plant. In case of failure of one 

HPIS train, the cladding temperature of the upper part of the fuel elements may rise slightly but 

soon it is returned to the safety values by the action of the accumulators. Only in the case of 

malfunction of both HPIS trains, the cladding temperature can rise above the steady state value. 

Even in this case, the accumulators injection returns the critical variables to normal values 

before the end of the transient.

After a first set of calculations, passive heat structures have been added to the model to 

simulate the heat accumulated in the vessel materials. After the sensitivity calculation performed 

(in which the total failure of HPIS scenario is repeated) the importance of simulating the passive 

heat structures has become clear. The heat accumulated by this material is released during the 

transient (at a rate that is the same order of magnitude than residual power during some 

hundreds of seconds), slowing the depressurization rate and causing a larger temperature 

excursion in the core than that calculated without those structures. Even in this case, however, 

accumulator injection is able to return the temperatures to normal values before the end of the 

transient.

Finally, the worst case of the scenarios studied (complete loss of the HPIS trains with passive 

heat structures) was recalculated with varying reactor coolant pumps stop times. This sensitivity 

analysis shows that imposing a pump trip simultaneously with the break is a conservative option

because the more the trip is delayed the milder are the cladding temperature peaks.
10-1

10 REFERENCES 

1. F. Reventós, L. Batet, C. Llopis, C. Pretel and I. Sol, Thermal-Hydraulic Analysis Tasks for 

ANAV NPPs in Support of Plant Operation and Control, Science and Technology of Nuclear 

Installations, Volume 2008.

2. The RELAP5 Code Development Team, RELAP5/Mod3 Code Manuals, NUREG/CR-5535, 

1995.

3. Pretel C., Batet L., Cuadra A., Machado A., San José G. de, Sol I., Reventós F., April 2000. 

Qualifying, validating and documenting a thermal-hydraulic code input deck. Workshop 

proceedings. Advanced thermal-hydraulic and neutronic codes: current and future 

applications, NEA/CSNI/R (2001) 2, Vol 2 pp.239-250.

4. F. Reventós, L. Batet, C. Pretel, M. Salvat, and I. Sol, Advanced qualification process for 

ANAV integral plant models, Nuclear Engineering and Design, vol. 237, no. 1, pp. 54–63, 

2007.

5. Central Nuclear de Ascó. IOE-E-1 Pérdida de refrigerante del reactor o secundario.

6. C.N. Vandellòs II, (in Spanish) Nota de cálculo del modelo de C.N. Vandellòs II con 

RELAP5/Mod3.2

7. Westinghouse; Model F steam generator termal-hydraulic data report (WTP-ENG-TN-81-

001)

8. Westinghouse Nuclear Española, (in Spanish) Descripción del SNGV de Westinghouse,

1983.

9. Westinghouse Energy Systems International; Precautions, limitations and setpoints. C.N. 

Vandellòs II