NUREG/IA-0257 

International 

Agreement Report

Simulation of PKL Loss of RHRS Experiment 

F2.2 Run 2 with RELAP5 and TRACE Codes – 

Application to a PWR NPP Model 

 

 Prepared by: 

 César Queral1

, Antonio Expósito1, and Alberto Concejal1

 

 1 Universidad Politécnica de Madrid 

 Alenza 4. 28003 Madrid 

 SPAIN 

 

A. Calvo, NRC Project Manager 

 Office of Nuclear Regulatory Research 

 U.S. Nuclear Regulatory Commission 

 Washington, DC 20555-0001 

Manuscript Completed: June 2011 

Date Published: September 2011

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-0257

Simulation of PKL loss of RHRS experiment 

F2.2 run2 with RELAP5 and TRACE codes.

Application to a PWR NPP model

Manuscript Completed: June 2011

Date Published: September 2011

Prepared by:

César Queral1

, Antonio Expósito1 and Alberto Concejal1

1

Universidad Politécnica de Madrid

Alenza 4. 28003 Madrid. 

Spain.

A. Calvo, NRC Project Manager

Prepared for:

Division of Systems Analysis 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

ABSTRACT

An analysis of PKL midloop test F2.2 has been performed with TRACE and RELAP5/MOD3

codes. This test, included within the OECD/PKL project, tries to analyze the phenomenology and different accident management actions after a loss of RHRS at midloop conditions

with primary side closed. The comparison between the results obtained with both codes

shows that, in general, the main phenomena are well reproduced. However, there are still

a few phenomena that are not well predicted, like pressurizer level. The good results obtained simulating another similar test (PKL test E3.1) enables to confirm that the modelling

methodology is adequate for this kind of transients. This modelling methodology has been

also applied to a TRACE model of a Spanish PWR, Westinghouse design, simulating a

similar transient to test F2.2 run2, obtaining similar results than in PKL models.

iii
iv
FOREWORD

Extensive knowledge and techniques have been produced and made available in the field

of thermal-hydraulic responses during reactor transients and accidents, and major system

computer codes have achieved a high degree of maturity through extensive qualification,

assessment and validation processes. Best-estimate analysis methods are increasingly used

in licensing, replacing the traditional conservative approaches. Such methods include an

assessment of the uncertainty of their results that must be taken into account when the

safety acceptance criteria for the licensing analysis are verified.

Traditional agreements between the Nuclear Regulatory Commission of the United States of

America (USNRC) and the Consejo de Seguridad Nuclear of Spain (CSN) in the area of nuclear safety research have given access to CSN to the NRC-developed best estimate thermalhydraulic codes RELAP5, TRAC-P, TRAC-B, and currently TRACE. These complex tools,

suitable state-of-the-art application of current two-phase flow fluid mechanics techniques to

light water nuclear power plants, allow a realistic representation and simulation of thermalhydraulic phenomena at normal and incidental operation of NPP. Owe to the huge required

resources, qualification of these codes have been performed through international cooperation programs. USNRC CAMP program (Code Applications and Maintenance Program)

represents the international framework for verification and validation of NRC TH codes,

allowing to: Share experience on code errors and inadequacies, cooperating in resolution

of deficiencies and maintaining a single, internationally recognized code version; Share user

experience on code scaling, applicability, and uncertainty studies; Share a well documented

code assessment data base; Share experience on full scale power plant safety-related analyses

performed with codes (analyses of operating reactors, advanced light water reactors, transients, risk-dominant sequences, and accident management and operator procedures-related

studies); Maintain and improve user expertise and guidelines for code applications.

Since 1984, when the first LOFT agreement was settled down, CSN has been promoting

coordinated joint efforts with Spanish organizations, such as UNESA (the association of

Spanish electric energy industry) as well as universities and engineering companies, in the

aim of assimilating, applying, improving and helping the international community in the

validation of these TH simulation codes, within different periods of the associated national

programs (e.g., CAMP-Espa ̃na). As a result of these actions, there is currently in Spain a

good collection of productive plant models as well as a good selection of national experts in

the application of TH simulation tools, with adequate TH knowledge and suitable experience

on their use.

Many experimental facilities have contributed to the today’s availability of a large thermalhydraulic database (both separated and integral effect tests). However there is continued

need for additional experimental work and code development and verification, in areas where

no emphasis have been made along the past. On the basis of the SESAR/FAP reports Nuclear

Safety Research in OECD Countries:Major Facilities and Programmes at Risk (SESAR/FAP,

2001) and its 2007 updated version Support Facilities for Existing and Advanced Reactors

(SFEAR) NEA/CSNI/R(2007)6, CSNI is promoting since 2001 several collaborative internav
tional actions in the area of experimental TH research. These reports presented some findings

and recommendations to the CSNI, to sustain an adequate level of research, identifying a

number of experimental facilities and programmes of potential interest for present or future

international collaboration within the safety community during the coming decade.

CSN, as Spanish representative in CSNI, is involved in some of these research activities,

helping in this international support of facilities and in the establishment of a large network of

international collaborations. In the TH framework, most of these actions are either covering

not enough investigated safety issues and phenomena (e.g., boron dilution, low power and

shutdown conditions), or enlarging code validation and qualification data bases incorporating

new information (e.g., multi-dimensional aspects, non-condensable gas effects). In particular,

CSN is currently participating in the PKL and ROSA programmes.

The PKL is an important integral test facility operated by of AREVA-NP in Erlangen

(Germany), and designed to investigate thermal-hydraulic response of a four-loop Siemens

designed PWR. Experiments performed during the PKL/OECD program have been focused

on the issues: Boron dilution events after small-break loss of coolant accidents; Loss of

residual heat removal during mid-loop operation (both with closed and open reactor coolant

system.

ROSA/LSTF of Japan Atomic Energy Research Institute (JAERI) is an integral test facility designed to simulate a 1100 MWe four-loop Westinghouse-type PWR, by two loops

at full-height and 1/48 volumetric scaling to better simulate thermal-hydraulic responses

in large-scale components. The ROSA/OECD project has investigated issues in thermalhydraulics analyses relevant to water reactor safety, focusing on the verification of models

and simulation methods for complex phenomena that can occur during reactor transients and

accidents such as: Temperature stratification and coolant mixing during ECCS coolant injection; Water hammer-like phenomena; ATWS; Natural circulation with super-heated steam;

Primary cooling through SG depressurization; Pressure vessel upper-head and bottom break

LOCA.

This overall CSN involvement in different international TH programmes has outlined the

scope of the new period of CAMP-Espa ̃na activities focused on: Analysis, simulation and

investigation of specific safety aspects of PKL/OECD and ROSA/OECD experiments; Analysis of applicability and/or extension of the results and knowledge acquired in these projects

to the safety, operation or availability of the Spanish nuclear power plants. Both objectives

are carried out by simulating experiments and plant application with the last available versions of NRC TH codes (RELAP5 and TRACE). A CAMP in-kind contribution is aimed as

end result of both types of studies.

Development of these activities, technically and financially supported by CSN, is being carried out by 5 different national research groups (Technical Universities of Madrid, Valencia

and Catalu ̃na). On the whole, CSN is seeking to assure and to maintain the capability of

the national groups with experience in the thermal hydraulics analysis of accidents of the

Spanish nuclear power plants.

Francisco Fern ́andez Moreno, Commissioner Consejo de Seguridad Nuclear (CSN)

vi
CONTENTS

Page

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

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

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

ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi

ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xiii

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

2 DESCRIPTION OF PKL FACILITY AND SELECTED TEST . . . . . . . 2-1

3 TRACE 5.0 AND RELAP5 MODELS OF PKL FACILITY . . . . . . . . . 3-1

4 COMPARISON RESULTS OF F2.2 RUN 2 TEST WITH TRACE AND RELAP5 CODES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

5 SIMULATION OF LOSS OF RHRS SEQUENCES AT MIDLOOP CONDITIONS IN ALMARAZ NPP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1

6 EXECUTION STATISTICS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1

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

Figures

Page

1 PKL facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

2 PKL vessel. TRACE model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1

3 PKL steam generators. TRACE model . . . . . . . . . . . . . . . . . . . . . . . . . 3-2

4 PKL pressurizer. TRACE model . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3

5 PKL hot leg and cold leg. TRACE model . . . . . . . . . . . . . . . . . . . . . . . 3-4

6 Reactor vessel pressure in the upper plenum. Test F2.2 . . . . . . . . . . . . . . . . 4-2

vii
7 Liquid temperature in the upper plenum of the reactor vessel. Test F2.2 . . . . . . 4-2

8 Pressurizer level. Test F2.2 run2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3

9 Vessel level. Test F2.2 run2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3

10 Steam generators pressures. Test F2.2 . . . . . . . . . . . . . . . . . . . . . . . . . 4-4

11 Steam generators water level. Test F2.2 . . . . . . . . . . . . . . . . . . . . . . . . 4-4

12 Heat transfer to steam generators and heat losses in primary side. Test F2.2. TRACE

code . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5

13 TRACE model of Almaraz NPP (SNAP mask with mean components) . . . . . . . 5-3

14 Nodalization of the TRACE model of 3D VESSEL . . . . . . . . . . . . . . . . . . 5-4

15 RCS pressure. Transients similar to PKL test F2.2 run 2 . . . . . . . . . . . . . . . 5-5

16 Collapsed vessel level. Transients similar to PKL test F2.2 run 2 . . . . . . . . . . . 5-5

17 Maximum rod temperature. Transients similar PKL test F2.2 run 2 . . . . . . . . . 5-6

18 Level of SG3. Transient equivalent to PKL test F2.2 run 2 . . . . . . . . . . . . . . 5-6

Tables

Page

1 Primary side initial conditions. Test PKL F2.2 run 2 . . . . . . . . . . . . . . . . . 2-3

2 Secondary side initial conditions. Test PKL F2.2 run 2 . . . . . . . . . . . . . . . . 2-3

3 Test PKL F2.2 run 2. Event sequence . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3

4 Heat transfer balance in primary side. PKL F2.2. TRACE and RELAP5 codes. . . 4-5

5 Simulations results for the loss of RHRS at midloop level and closed RCS. Almaraz NPP

model and experimental data from PKL test F2.2 run2 . . . . . . . . . . . . . . . . 5-3

viii
EXECUTIVE SUMMARY

The PKL III test facility models a typical 4 loop 1300 MWe pressurized water reactor of

Siemens/KWU design. As a part of the PKL/OECD program the PKL III-F2 test series have

been performed. This series includes several tests that have been carried out to investigate

the thermal hydraulic phenomenology that entails the loss of RHRS system at mid-loop

conditions (low power - shutdown plant state) and the different accident management actions

that could be implemented by operators during the transient.

In test F2.2 run2, the primary inventory was fixed at midloop level and thus, the capability of

the reflux-condensation (including the influence of noncondensables gases) as a mechanism

of cooling was investigated. In this test the RCS is closed and two steam generators are

available, both of them operable. The subsequent operation of the other two SGs, which

were empty of water and filled with nitrogen at the beginning of the transient, is used to

recover a decay heat sink.

In this report, a post-test analysis of PKL III-F2.2 run2 using RELAP5/MOD3 and TRACE

codes is presented. A description of the model inputs is given, and the comparison of measured and calculated results is discussed. Simulation results of equivalent transient conditions

implemented in a Spanish PWR Westinghouse design (3-loop) plant model are also discussed

in the report.

The purpose of this analysis is to contribute in the validation of TRACE and RELAP5

codes and its ability to properly simulate shutdown conditions. The main findings of the

comparison of RELAP5/Mod 3.2 and TRACE 5.0 results with the PKL III F2.2 experiment

are:

• Two phase reflux - condensation cooling mechanism with non-condensable gases in the

primary side is well reproduced. The net heat balance obtained shows good agreement

with experimental data.

• A wrong primary mass distribution has been obtained, with a very large water level

in the PZR. It could be interesting to check the condensation and offtake correlations

for this and other geometries (pressurizer surge line connection with the hot leg and

break in the vessel head) in both codes.

• The main phenomena of AFW injection phase have been also well reproduced.

In addition, the application to a mid-loop transient in a PWR Westinghouse shows similar

trends that the results obtained in PKL test.

ix
x
ACKNOWLEDGMENTS

This (paper/work) contains findings that were produced within the OECD-NEA (PKL/ROSA)

Project. The authors are grateful to the Management Board of the (PKL/ROSA) Project

for their consent to this publication, and thank the Spanish Nuclear Regulatory Body (CSN)

for the technical and financial support under the agreement STN/1388/05/748.

xi
xii
ABBREVIATIONS

CAMP Code Applications and Maintenance Program

CTC Consolidated Thermohydraulic Code

CCFL Counter-Current Flow Limit

CSN Consejo de Seguridad Nuclear (Spanish Nuclear Regulatory Commission)

PWR Pressurized Water Reactor

PZR Pressurizer

RCS Reactor Coolant System

RHRS Residual Heat Removal System

SETH SESAR Thermal Hydraulics

SESAR Senior Group of Experts on Nuclear Safety Research

SG Steam Generator

xiii
1 INTRODUCTION

The PKL III test facility simulates a typical 4 loop 1300 MWe pressurized water reactor of

Siemens/KWU design. As a part of the PKL/OECD program the PKL III-F2 test series have

been performed. This series includes several tests that have been performed to investigate the

thermohydraulic phenomenology that entails the loss of RHRS system at midloop conditions

(low power - shutdown plant state) and the different accident management actions that could

be implemented by operators during the transient.

In test F2.2 run2, the primary inventory was at midloop level and thus, the capability of

the reflux-condensation, including the influence of noncondensables gases, as a mechanism

of cooling was investigated. In this test the RCS is closed and two steam generators are

available, both of them operable. The subsequent operation of the other two SGs, which

were empty of water and filled with nitrogen at the beginning of the transient, is used to

recover a decay heat sink.

In this report, a post-test analysis of PKL III-F2.2 using RELAP5/MOD3 and TRACE codes

is presented. A description of the model inputs is given, and the comparison of measured

and calculated results is discussed. Simulation results of equivalent transient conditions

implemented in the model of a W-design Spanish 3-loop plant, are also discussed in the

report.

The purpose of this analysis is to contribute in the validation of TRACE code and its ability

to properly simulate shutdown conditions.

1-1
2 DESCRIPTION OF PKL FACILITY AND

SELECTED TEST

The PKL test facility is a large scaled-down model of a four loop PWR-KWU design with

1300 MWe (Reference plant: Philippsburg 2 NPP), Figure 1. Its main characteristics are,

• Elevations are scaled 1:1 while the volumes, power and mass flows are scaled 1:145.

• The reactor core is modeled by a bundle of 314 electrically heated rods with a maximum

power of 2.5 MW (10% of rated power).

• Maximum operating pressure is 45 bar.

• Each steam generator has 30 U-tubes of original size and material.

Figure 1: PKL facility

This facility allows the detailed analysis of several accident scenarios like: SBLOCA, LBLOCA

or loss of RHRS at midloop conditions.

Run 2 of PKL test F2.2 is a sequence of loss of RHRS at midloop conditions. At the beginning

of the test, the facility is shut down and operating at midloop conditions, Table 1. Steam

2-1
generators 1 and 2 (SG1 and SG2) are full of water, while steam generators 3 and 4 (SG3

and SG4) are full of nitrogen, Table 2. SG1 and SG2 are pressure controlled with a pressure

setpoint of 2 bar. The total core power is 219 kW, which is equivalent to the 0.7% of the

nominal power, in order to simulate the decay heat, plus 38 kW to take into account the

heat losses of the facility. The test F2.2 has four stages, Table 3:

1. The transient starts with the loss of the RHRS. The pressure and temperature of

the primary system start rising quickly after saturation temperature is reached. The

steam generated in the core condensates at SG1 and SG2 U-tubes and mainly comes

back to the vessel. This heat transfer mechanism between the primary and secondary

side is called reflux and condensation cooling. Once the sum of heat transferred into

the secondary side and primary heat losses balance core power, primary pressure and

temperatures reach quasi-stationary values. During this phase of the transient the temperature and pressure in both circuits keep growing until the pressure in the secondary

side reaches 2 bar.

2. When SG1/SG2 reach 2 bar begins the second phase. From this point in time on, steam

pressure control system maintains pressure at 2bar in these two SGs; SG level control

system also starts to feed both SG. An additional heat sink, implemented through

the later actuation of SG3/SG4 level control system, feeds these SGs until the level

setpoint is reached, and through subsequent actuation of SG3/SG4 pressure control

system, controlling also in 2 bar, allows to reach a new and lower quasi-stationary

primary pressure.

3. Two similar injection sequences, using the fuel pool cooling system (FPCS), were initiated for makeup of reactor coolant system inventory. This emergency measure allows

to recover primary inventory.

4. After last FPCS injection, RHRS becomes available, and starts to cool down RCS.

Only the first and second phases are simulated in this work because the main interest is to

analyze reflux-condensation mechanism and compare it with similar conditions in a realistic

case in an actual NPP.

2-2
Power 219kW = 0.7% + 38kW

Total mass inventory 1300 kg

PZR mass inventory 39 kg

PZR level 1.0 m

PZR temperature T ≃ 56 C

Core outlet temperature T ≃ 61, 5 C

Primary pressure P = 0, 95 · 105 P a

Table 1: Primary side initial conditions. Test PKL F2.2 run 2

SG1 SG2 SG3 SG4

SG level 12.2 m 12.2 m 0.0 m 0.0 m

SG temperature 33 − 47 C 33 − 47 C 33 − 47 C 33 − 47 C

FW temperature 26.0 C 26.0 C 26.0 C 26.0 C

Operation YES YES YES YES

Table 2: Secondary side initial conditions. Test PKL F2.2 run 2

Time Event

0 s Shutdown of RHRS

8961 s SG1 pressure control activated

10447 s SG2 pressure control activated

10981 s begin of feedwater feeding in SG1

11825 s begin of feedwater feeding in SG2

18543 s begin of feedwater feeding in SG3

22783 s begin of feedwater feeding in SG4

30563 s Begin of FPCS injections

37818 s RHRS started

39547 s end of the test

Table 3: Test PKL F2.2 run 2. Event sequence

2-3
3 TRACE 5.0 AND RELAP5 MODELS OF PKL

FACILITY

RELAP5 model of PKL is based on an input deck provided by AREVA, operator of the

facility, to the participants in the OECD/PKL project. This RELAP5 model was checked

and several parameters adjusted in order to improve the results. The TRACE model of PKL

was derived from the previous RELAP5 model. This model has 330 thermo-hydraulic cells,

368 SIGNAL VARIABLE, 341 CONTROL BLOCK and 15 TRIP. The vessel is modeled

with both downcomer pipes as in PKL facility, Figure 2. Each hot leg is modeled with a

4 cells PIPE in order to avoid convergence problems in the calculations (the models with

more cells showed instability problems). The primary side of the steam generators are

modeled by three PIPE of different heights in order to simulate better the heat transfer

from primary to secondary side as well as RCS flow regimes (e.g., onset and interruption of

natural circulation), Figure 3. In both models, heights and volumes of each component were

checked against the PKL data.

T−130−side

T−130−main

2

2

1

1

2

3

4

5

6

7

1

2

3

P−110

8

P−130

P−121

Jun−130

Jun−110

P−120

P−104

Jun−104

1

2

3

4

5

P−105

P−139

1

Jun−103 Jun−120 Jun−121

1

2

3

Jun−150

P−150

Pl−160

Jun−109

P−109 Jun−160

P−108

P−99

P−100

Jun−99

Jun−140

Jun−108

Jun−139 P−140

Jun−102

Jun−131

Jun−105

Jun−209

Jun−309

Jun−409

Jun−009

Jun−265

Jun−365

Jun−465

Jun−065

5

1

2

3

4

2.6992

3.9742

4.5992

5.2492

5.7392

3.3242

7.5311

9.0325

7.8804

1.75255

0.5425

1.75255

3.47955

6.2827

5.0005

7.1000

6.425

3.47955

5.0005

7.45025

0.1417

6.425

6.76025

2.04925

0.7142

3.34925

5.08925

7.0481

0.5425

9.7692

8.6437

7.9088

1.4117

2.0492

Figure 2: PKL vessel. TRACE model

3-1
2

P−600 1

2

111213141516171819

P−237P−235P−236

J−229 J−231 J−23

J−225

121110

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1110

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123

J−611

2

P−630

Jun−620

Jun−625

P−601

4

1

P−610

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12

9

1110

T−620

P−611

13 1

5

SJ−623

Jun−623

Jun−615

Jun−621

Jun−600

J−230 P−230

20

Pl−680

P−625

P−210

12345

6789

10

Jun−610

23.51425

11.3868

23.06925

17.5818

22.5318

21.94680

21.41055

19.646815.5168

13.4518

23.94475

20.92305

0.6912

1.112950.651.001.001.001.001.000.9671.001.000.48791.211

12.47

0.35

10.35425

20.0898319.1877518.598217.8542510.5292522.8242522.211821.861811.0292515.8542516.8542514.8542513.8542512.8542511.85425

Figure 3: PKL steam generators. TRACE model

3-2
2

1

3

2

18

5

P−525

T−518

Jun−520

Jun−525

P−521 P−522

15

V−565 P−520

Jun−520

B−566

Jun−318 P−312

3

P−510

6

1

17

F−560 F−562

16

P−519

Jun−519

Jun−518

Jun−510

13

12

11

10

9

8

6

5

4

7

4

14

19.3548

6.888

7.400

6.660

4.920

8.400

6.550

7.489

8.680

9.330

10.630

11.280

11.930

12.580

13.230

14.540

15.540

15.840

17.140

17.790

18.440

19.350

8.030

13.880

15.190

16.490

7.173

6.40325

8.680

9.330

10.630

11.280

11.930

12.580

13.230

14.540

15.540

15.840

17.140

17.790

18.440

19.350

7.173

8.030

13.880

15.190

16.490

7.489

8.680

9.330

10.630

11.280

11.930

12.580

13.230

14.540

15.540

15.840

17.140

17.790

18.440

19.350

7.173

8.030

13.880

15.190

16.490

7.489

19.3548

7.80375

1m

Figure 4: PKL pressurizer. TRACE model

3-3
Pl−100


J−241J−240J−239

 U−tubes

J−241J−240J−239

 U−tubes

Pl−140

4

3

2 1 Jun−209

Jun−225

P−230

P−247

P−240

Jun−246

P−248

Jun−245

1 2 3 4 5

Jun−247

P−249

5

4

3

2

1

6

7

8

P−250

Jun−249

Jun−250

Pp−260

Jun−260

Jun−265

P−265

7.40175

9.95288

6.67813

5.81538

4.28875

10.26125

4.05625

8.98775

7.61175

9.51413

9.07538

10.26125

8.63663

8.32225

7.66838

7.45025

6.85625

5.95625

4.95625

4.25625

7.30475

4.95263

3

4

5

V−215

P−216

Jun−216

Jun−210

1

2

P−210

Figure 5: PKL hot leg and cold leg. TRACE model

3-4
4 COMPARISON RESULTS OF F2.2 RUN 2 TEST

WITH TRACE AND RELAP5 CODES

In order to describe the results, the main thermal hydraulic variables of the simulation are

clustered in two sets, one related with primary side and the other to the secondary side. The

comparison of experimental and simulated data show the following results:

• Primary pressure and temperature show a good agreement with experimental data

reaching slightly different equilibrium values with two SG filled of water, Figures 6 and

7. However, the simulated and experimental values are quite similar in the last stage

of the transient, when there is water in all SG. These results indicate that heat transfer

between primary side and steam generators have been well reproduced, mainly in the

last stage of the transient.

• The mass distribution in primary side is not completely well reproduced. The main

reason is that a large amount of water goes to PZR in both simulations, and they are

larger than the experimental data, Figure 8. With respect to collapsed vessel level, it

is observed that the results of RELAP5 are better than TRACE code ones, Figure 9.

• The heat transfer from primary to secondary side causes a pressure increase in all steam

generators when they are filled with water, Figures 10 and 11. The level in all SG is

well reproduced by both codes, Figure 11. However, the simulation with RELAP5

shows a peak in the secondary pressure at the beginning of feedwater injection in SG3

and SG4, Figure 10. Apart from this, the secondary pressures are well reproduced by

TRACE and RELAP5 codes.

• The analysis of heat transfer in this transient shows an interesting behavior, Figure 12

and Table 4. It is observed that both SGs filled of water have the same heat transfer

capability in the first stage of the test. When SG3 and SG4 are partially filled with

water the heat transfer from primary to secondary grows up and primary pressure

decreases. The comparison of heat transfer results, Table 4, shows that the SGs that

are partially filled with water have the same heat transfer capability than the SGs with

higher level. Also it is observed that the primary heat losses are higher in RELAP5

simulation.

In general, test F2.2 run 2 is well reproduced. However, mass distribution problems are

observed, mainly related with the high water level obtained for the pressurizer.

4-1
0 10000 20000 30000

Time (s)

0

2e+05

4e+05

6e+05

8e+05

pressure (Pa)

Experimental

TRACE

R5

Figure 6: Reactor vessel pressure in the upper plenum. Test F2.2

0 5000 10000 15000 20000 25000 30000

Time (s)

300

350

400

450

liquid temperature (K)

Experimental

TRACE

R5

Figure 7: Liquid temperature in the upper plenum of the reactor vessel. Test F2.2

4-2
0 5000 10000 15000 20000 25000

Time (s)

0

5

10

15

Level (m)

Experimental

TRACE PZR level

R5 PZR level

Figure 8: Pressurizer level. Test F2.2 run2

0 10000 20000 30000

Time (s)

3

4

5

6

7

8

9

10

Level (m)

Experimental

TRACE Hot leg

R5 vessel level in

Figure 9: Vessel level. Test F2.2 run2

4-3
Experimental

R5 SG1

TRACE SG1

TRACE SG2

R5 SG2

0 10000 20000 30000

Time (s)

1e+05

1.5e+05

2e+05

2.5e+05

3e+05

pressure (Pa)

Experimental

TRACE SG3

R5 SG3

R5

TRACE

Figure 10: Steam generators pressures. Test F2.2

0 10000 20000 30000

Time (s)

0

5

10

15

20

Level (m)

Exp SG1

Exp SG2

Exp SG3

Exp SG4

TRACE SG1

TRACE SG2

TRACE SG3

TRACE SG4

R5 SG1

R5 SG2

R5 SG3

R5 SG4

Figure 11: Steam generators water level. Test F2.2

4-4
Figure 12: Heat transfer to steam generators and heat losses in primary side. Test F2.2.

TRACE code

TRACE RELAP5

17500 s 22000 27000 17500 s 22000 27000

SG 1 43.0% 29.3% 22.5% 42.0% 28.3% 21.5%

SG 2 43.0% 29.3% 22.5% 42.0% 28.3% 21.5%

SG 3 2.0% 29.3% 22.5% 1% 28.3% 21.5%

SG 4 2.0% 2.0% 22.5% 1% 1% 21.5%

Primary heat losses 10.0% 10.0% 10.0% 14.0% 14.0% 14.0%

Total heat transfer 100% 100.0% 100.0% 100% 100.0% 100.0%

Total core power 100.0% 100.0% 100.0% 100% 100.0% 100.0%

Table 4: Heat transfer balance in primary side. PKL F2.2. TRACE and RELAP5 codes.

4-5
5 SIMULATION OF LOSS OF RHRS SEQUENCES

AT MIDLOOP CONDITIONS IN ALMARAZ

NPP

The development of the Almaraz NPP model for TRACE code has been framed within several

national and international projects (CAMP, CTC, SETH and PKL-OECD) sponsored in

Spain by the Nuclear Regulatory Body, Consejo de Seguridad Nuclear (CSN) and the electric

energy industry of Spain (UNESA). In these projects, one of the most important objectives

is the maintenance and development of the Spanish NPP models with the thermal-hydraulic

codes that have been sponsored by NRC, such as RELAP5, TRAC-P, TRAC-M and TRACE.

The capability of the TRACE code in simulating low power and shutdown sequences has

been confirmed in the simulation of some PKL experiments by several groups, as it has been

showed in previous section.

Almaraz NPP is a three loop PWR-W. The TRACE model of Almaraz NPP, Figure 13, is

made up of 252 thermal-hydraulic components, 1 VESSEL, 52 PIPE, 71 TEE, 41 VALVE, 3

PUMP, 20 FILL, 27 BREAK, 36 HEAT STRUCTURE and 1 POWER component and also

by logical and control systems signals, 685 SIGNAL VARIABLE, 1532 CONTROL BLOCK

and 47 TRIP.

Regarding the primary circuit, the following components have been modeled,

• Reactor vessel, modeled by a 3D VESSEL component which includes the core region,

guide tubes, support columns, core bypass, and the bypass to the vessel head via

downcomer and via guide tubes, Figure 14.

• The three loops, each one composed of pump and steam generator, and pressurizer in

loop 2 (containing heaters, relief and safety valves and pressurizer spray system).

• Chemical and volumetric control system.

• Safety injection system and accumulators.

With reference to the secondary circuit, the following components have been modeled,

• Normal feedwater and auxiliary feedwater systems of the steam generators.

• The steam lines up to the turbine stop valves, with the relief, safety and isolating

valves, and the steam dump with the eight valves.

The shutdown model of Almaraz NPP has been obtained from the model at full power

operation. The following changes have been performed:

• every automatic control of the plant has been deactivated, except the level control of

steam generator.

5-1
• every thermal-hydraulic volume of the model has been initialized to values of LPS

conditions.

• a vessel level control has been implemented in order to reach the desired level (midloop

level, vessel flange level, level in pressurizer).

• a fine nodalization in vessel has been implemented in order to reach midloop conditions

in vessel as well as in cold and hot legs.

• the CCFL model has been taken into account in surge line and U-tubes.

• the offtake model on the surge line connection to the pressurizer has been considered.

An application case of the PKL test F2.2 run 2 for Almaraz NPP has been performed starting

from the following conditions:

• Midloop level (mass inventory of almost 73000 kg)

• Closed primary. Initial pressure 1 bar, initial temperature 333.15K

• Decay heat: 11 MW of thermal power

• Availability of steam generators: 2 steam generators full of water and 1 full of air

(2SG).

• Secondary pressure: 1 bar

• Auxiliary Feedwater available for SG3 when equilibrium primary pressure conditions

are reached (13000 s.).

The case with 3 steam generators full of water (3SG) it is also simulated in order to check

the final steady state in the application case described above.

The results obtained show that the transient behavior is similar in Almaraz NPP model and

PKL experimental data, Figures 15 to 17. There are two stages in the transient:

• In the first part of the transient the pressure equilibrium condition is reached after

13000 s, similar to PKL F2.2 run 2 (T= 15000 s).

• In the second part (T>13000 s) the auxiliary feedwater is available for SG3 and a

new stationary state is reached after ∼ 500 s (T= 13500 s), Figures 15 to 17. The

pressure decreases from 5.5 bars to 3.5 bars, obtaining better cooling conditions (lower

temperatures) and lower pressures.

Concerning the gravity feed from the RWST, the maximum pressure for gravity feed is

conditioned by the specific plant configuration, but this pressure is generally between

2 and 3 bar. Therefore, this result also indicates that if AFW is injected soon in the

empty SG it is possible to increase the available time to inject with gravity feed from

RWST.

5-2
So, it is observed that the Almaraz NPP results are similar to PKL F 2.2 run 2. In fact, the

comparison of the equilibrium pressure values of Almaraz NPP model with respect to PKL

data also show a similar order of magnitude, Table 5.

Case Maximum Pressure (bar) Time to 2 bar (min.) Time to 3 bar (min.)

2/3 SG Almaraz 5.6 33 62

3/3 SG Almaraz 3.8 45 85

2/4 SG PKL 4 85 150

3/4 SG PKL 3 – –

4/4 SG PKL 2.7 – –

Table 5: Simulations results for the loss of RHRS at midloop level and closed RCS. Almaraz

NPP model and experimental data from PKL test F2.2 run2

Figure 13: TRACE model of Almaraz NPP (SNAP mask with mean components)

5-3
 

Z0

Z1

Z2

Z3

Z4

Z5

Z6

Z7

Z8

Z9

Z10

Z11

Z12

Z13

Z14

Z15

Z16

Z17

Z18

1

2

3

4

6

5

7

8

9

10

11

12

13

14

15

16

17

18

Z19

Z20

Z21

Z22

Z23

Z24

19

20

21

22

23

24

R2 R3 R4 R1

Hot leg conexion

1 m

4

5 6

8

11

7

10

9

12

2 3

1

R2

R1

R3

R4

DOWNCOMER

CORE

4

5 6

8

11

7

10

9

12

2 3

1

R2

R1

R3

R4

Support column

Guide tube

DOWNCOMER

CORE

Core Bypass

4

5 6

8

11

7

10

9

12

2 3

1

R2

R1

R3

R4

BARREL

CORE

hot leg

cold leg hot leg

cold leg

cold leg

hot leg

Support columns

Cold leg conexion

Guide tubes

Figure 14: Nodalization of the TRACE model of 3D VESSEL

5-4
0 1500 3000 4500 6000 7500 9000 10500 12000 13500 15000

Time (s)

0

1e+05

2e+05

3e+05

4e+05

5e+05

6e+05

Pressure (Pa)

Primary pressure. 3 SG

Primary pressure. 1 SG

RWST gravity feed (3 bar)

RWST gravity feed (2 bar)

Figure 15: RCS pressure. Transients similar to PKL test F2.2 run 2

0 5000 10000 15000

Time (s)

5

6

7

8

9

cb 6020

Vessel level. 3SG

Vessel level. 2SG

Active core level

Figure 16: Collapsed vessel level. Transients similar to PKL test F2.2 run 2

5-5
0 5000 10000 15000

Time (s)

300

400

500

600

700

max hot rod temp (K)

trhmax 3 SG

trhmax 2 SG

Figure 17: Maximum rod temperature. Transients similar PKL test F2.2 run 2

13000 13500 14000 14500 15000 15500

Time (s)

0

2

4

6

8

Level (m)

Figure 18: Level of SG3. Transient equivalent to PKL test F2.2 run 2

5-6
6 EXECUTION STATISTICS

The simulations have been run in Pentium IV 3.4 MHz under Windows XP and AMD

Opteron Dual Core Processors 180 & 1222 under Debian, both with 32 and 64 bits precompiled executables provided by NRC. No significant differences where found between runs

executed in Windows and Debian systems, and between 32 and 64 bits code versions.

6-1
7 CONCLUSIONS

The PKL III test facility simulates a typical 1300 MWe pressurized water reactor of Siemens

/ KWU design. In test F2.2 run2, the primary inventory was at midloop level and thus, the

influence of noncondensables gases on reflux-condensation cooling mechanism was investigated. In this test was also analyzed the impact of injecting AFW in the two SG initially

empty.

In this report, a post-test analysis of PKL test F2.2 run 2 using RELAP5/MOD3 and TRACE

codes has been presented. A description of the model inputs are given, and the comparison

of measured and calculated results is discussed.

The main findings of the comparison of RELAP5/Mod 3.2 and TRACE 5.0 results with the

PKL III F2.2 experiment are:

• Two phase reflux - condensation cooling mechanism with non-condensables gases in

the primary side is well reproduced, obtained a net heat balance quite accurate when

compared with experimental data.

• A wrong primary mass distribution has been obtained, with a very large water level

in the PZR. Probably, it could be interesting to check the heat transfer condensation

correlations and offtake correlations for this and other geometries (pressurizer surge

line connection with the hot leg and break in the vessel head).

• The main phenomena of AFW injection phase have been also well reproduced.

With respect to the application to a midloop transient in a PWR Westinghouse similar

results that in PKL test F2.2 run2 have been obtained.

7-1