NUREG/IA-0256 

International 

Agreement Report

Simulation of PKL Loss of RHRS Experiment E3.1 

with RELAP5 and TRACE Codes – Application to 

a PWR NPP Model 

 

 Prepared by: 

 César Queral1

, Alberto Concejal1, Antonio Expósito1, and Issac González1, 

 

 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-0256 

International 

Agreement Report

Simulation of PKL Loss of RHRS Experiment E3.1 

with RELAP5 and TRACE Codes – Application to 

a PWR NPP Model 

 

 Prepared by: 

 César Queral1

, Alberto Concejal1, Antonio Expósito1, and Issac González1, 

 

 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

ABSTRACT

An analysis of PKL midloop test E3,1 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 F2.2 run2) 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 E3.1, obtaining similar results than in PKL models.

iii

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

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 E3.1 TEST WITH TRACE AND RELAP5

CODES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

5 SIMULATION OF LOSS OF RHRS SEQUENCES AT MIDLOOP CONDITIONS FOR 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 Pressurizer pressure. Test E3.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4

vii
7 Core exit liquid temperature. Test E3.1 . . . . . . . . . . . . . . . . . . . . . . . . . 4-4

8 Steam generator 1 vapor mass flow. Test E3.1 . . . . . . . . . . . . . . . . . . . . . 4-5

9 Steam generators pressure. Test E3.1 . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5

10 Heat transfer to steam generators and heat losses in primary side. Test E3.1. TRACE

code . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6

11 Pressurizer water level. Test E3.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7

12 Vessel collapsed water level. Test E3.1 . . . . . . . . . . . . . . . . . . . . . . . . . 4-7

13 U-tubes water level, steam generator 1. Test E-3.1 . . . . . . . . . . . . . . . . . . . 4-8

14 U-tubes water level, steam generator 2. Test E-3.1 . . . . . . . . . . . . . . . . . . . 4-8

15 Nodalization of the TRACE model of Almaraz NPP . . . . . . . . . . . . . . . . . . 5-3

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

17 RCS pressure. Transients similar to PKL E3.1 . . . . . . . . . . . . . . . . . . . . . 5-5

18 U-tubes level of SG operating. Transients similar to PKL E3.1 . . . . . . . . . . . . 5-5

19 Vessel collapsed water level. Transients similar to PKL E3.1 . . . . . . . . . . . . . 5-6

20 Execution time. PKL test E3.1. TRACE model . . . . . . . . . . . . . . . . . . . . 6-1

Tables

Page

1 Primary side initial conditions. Test PKL E3.1 . . . . . . . . . . . . . . . . . . . . . 2-2

2 Secondary side initial conditions. Test PKL E3.1 . . . . . . . . . . . . . . . . . . . 2-3

3 Test PKL E3.1. Event sequence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3

4 Initial conditions in test PKL E3.1. Steady state values obtained with RELAP5 and

TRACE codes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

5 Heat losses in TRACE and RELAP5 models. Temperature 373.74 K, loops massflow

8.5 kg/s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2

6 Adjustment of pressure drops in RELAP5 model for different mass flows. . . . . . . 4-2

7 Equilibrium heat transfer balances in the primary side before and after transient phase

two for both codes. PKL E3.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6

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

model and experimental data from PKL test E3.1 . . . . . . . . . . . . . . . . . . . 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 OECD/PKL program, the PKL III E 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 E3.1 test, the primary inventory was fixed at mid-loop level and thus, the capability of

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

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

available, one of them operable.

In this report, a post-test analysis of PKL III-E3.1 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 E3.1 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.

• All participants of the E3.1 benchmark obtained a wrong primary mass distribution,

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.

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

trends that the results obtained in PKL test.

ix

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

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. Within the PKL III-E3 test series, two tests were performed to

investigate the thermohydraulic phenomenology that entails the loss of RHRS system at

midloop conditions (low power - shutdown plant state).

In test E3.1, the primary inventory was at midloop level (∼ 58%) and thus, the influence of

noncondensables on two-phase cooling was investigated. In this test was also analyzed the

capability of the reflux-condensation as a mechanism of cooling in a closed RCS considering

the availability of two steam generators, one of them operable.

In this report, a post-test analysis of PKL III-E3.1 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 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.

PKL test E3.1 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 generators

2-1
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 is the only one which is pressure controlled with a pressure

setpoint of 2 bar while SG2 is isolated. The total core power is 217 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 E3.1 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 reaches 2 bar begins the second phase.

From this point in time on, the feedwater system in SG1 starts the feeding and the

steam pressure control system maintains pressure at 2bar in SG1 while SG2 is isolated

from secondary system and its pressure goes on growing. From this instant on, both

steam generators have different pressures which produces a inventory displacement in

the primary side from SG2 towards SG1, leading to an increase of heat transfer in SG1

and a decrease in SG2.

3. After of several hours (34000 s) begins the third stage of the test, in which there are

five accumulators injections, first to fourth injections are in cold leg and the last one

is in hot leg 1.

4. Finally, after the last accumulators injection, RHRS becomes available, and starts to

cool down RCS.

Power 217kW = 0.7% (residual heat) +38kW (heat losses)

Total mass inventory 1300 kg (≃ 58%)

PZR mass inventory 39 kg

PZR level 1.04 m

PZR liquid temperature T ≃ 56 oC

Core outlet temperature T ≃ 61 oC

Primary pressure P = 105 P a

Table 1: Primary side initial conditions. Test PKL E3.1

2-2
SG1 SG2 SG3 SG4

SG level (m) 12.2 12.2 0.0 0.0

SG temperature (oC) 66.0 66.0 33.0 33.0

FW temperature (oC) 26.0 26.0 26.0 26.0

Operation YES NO NO NO

Table 2: Secondary side initial conditions. Test PKL E3.1

Time (s) Event

0 Begin of transient

1914 Shutdown of RHRS

10139 Secondary side pressure has increased to 2 bar.

Start of secondary side pressure control at 2 bar in SG1.

35009 Start of ACC injections

Table 3: Test PKL E3.1. 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

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8

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5

1

2

3

4

2.6992

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4.5992

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7.5311

9.0325

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3.47955

6.2827

5.0005

7.1000

6.425

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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

1 P−600

2

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P−237P−235P−236

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23.51425

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13.4518

23.94475

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1.2110.9671.11295

0.6912

0.48791.001.001.001.001.001.001.000.65

12.47

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1.001.001.001.001.001.001.00

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22.8242522.211821.861820.0898319.1877518.598217.8542516.8542515.8542514.8542513.8542512.8542511.8542511.0292510.52925

10.35425 10.26125

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

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13

12

11

10

9

8

6

5

4

7

4

14

19.3548

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4.920

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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 E3.1 TEST WITH

TRACE AND RELAP5 CODES

In a first step, a vessel level control was included in order to obtain midloop level. Later,

initial conditions were adjusted obtaining a good agreement with experimental data, Table

4. RELAP5 model was adjusted in four steps:

• First, the mass of heat structures was checked and corrected where it was necessary.

• Second, heat losses were adjusted in primary and secondary side, Table 5.

• Third, friction factors were adjusted for four different mass flows, Table 6, and the

test E3.1 was simulated with this four groups of friction factors. The best result was

obtained with the friction factors corresponding to the adjustment at G= 12.5 kg/s.

• Fourth, the best simulation was analyzed in detail and it was observed that the agreement of primary pressure an secondary pressures was good enough.

After this analysis, a new model was also developed for TRACE code in order to check if

this code was also able to reproduced this transient. The comparison of TRACE results with

respect to RELAP5 model and the experimental data is showed in this section.

Component Initial conditions

E3.1 RELAP5 TRACE

Vessel

Upper plenum pressure (Pa) 1.05E+05 1.05E+05 1.05E+05

Core exit temperature (K) 333.15 333.15 332.5

Water level midloop midloop midloop

Pressurizer

Upper plenum pressure (Pa) 1.03E+05 1.03E+05 1.05E+05

Water level (m) 1.04 1.04 1.046

Stem generators (SG1 and SG2)

Secondary side pressure (Pa) 1.05E+05 1.05E+05 1.05E+05

Secondary side temperature (K) 339.15 339.15 338.5

Water level (m) 12.2 12.2 12.2

Table 4: Initial conditions in test PKL E3.1. Steady state values obtained with RELAP5

and TRACE codes.

4-1
Component Heat losses (kW)

Experimental TRACE RELAP5

Downcomer vessel 1.33 1.24 1.33

Downcomer pipes 1.61 1.63 1.61

Lower plenum 1.61 1.60 1.61

Core region 5.35 5.33 5.34

Upper plenum 1.60 1.63 1.59

Upper head 0.28 0.27 0.28

Hot legs 1.51 1.20 1.51

RCP and cold legs 7.26 7.01 7.22

Loop seal 2.11 2.04 2.11

Surge line 0.12 0.12 0.13

Pressurizer 0.79 0.80 0.78

Primary 23.57 23.18 23.53

Secondary 18.43 18.54 18.39

Total 42.00 41.72 41.91

Table 5: Heat losses in TRACE and RELAP5 models. Temperature 373.74 K, loops massflow

8.5 kg/s

Component Pressure drop - Exp. / RELAP5 (Pa)

G = 1.22 kg/s G = 2.4 kg/s G = 12.5 kg/s G = 25 kg/s

Total Vessel 500 / 499 1420 / 1420 27050 / 27057 101150 / 101130

Hot Leg -51 /-32 -100 / -100 -1458 / -1460 -6660 / -6676

Steam Generator 393 / 387 1484 / 1463 25750 / 25750 90910 / 90897

Loop seal 170 / 169 1020 / 1020 16000 / 16049 60000 / 60040

Cold Leg 1030 / 1029 4137 / 4126 3650 / 3665 14285 / 14293

Table 6: Adjustment of pressure drops in RELAP5 model for different mass flows.

4-2
The analysis shown in this report corresponds to the results obtained as part of our participation in an international benchmark with PKL test E3.1 in the framework of PKL/OECD

project. In this benchmark the simulations were performed only until accumulators injections, from t = 0 s to t = 30000 s of the experiment and there is not attempt to compare

the phase of the accumulators injections. The reason is that one of the main phenomena of

interest was the reflux cooling in the steam generators.

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 the same equilibrium values, Figures 6 and 7. These results indicate that

heat transfer between primary side and steam generators have been well reproduced.

This hypothesis can be confirmed with the good results of steam mass flow in SG1,

Figure 8.

• 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 this

quantity is lower in the experimental data, Figure 11. With respect to the vessel

collapsed water level is well reproduced by RELAP5 but not by TRACE code, Figure

12. This difference could be explained taking into account the difference of the PZR

level obtained by both codes, Figure 11, and the section areas in PZR and vessel.

Finally, the levels in SG1 and SG2 are quite well reproduced by TRACE code but not

by RELAP5 code, Figures 13 and 14.

So, it is observed that some levels are better predicted by RELAP5 while others are

better reproduced by TRACE code.

• The heat transfer from primary to secondary side causes initially a pressure increase

in both steam generators filled with water, Figure 9. In this case, it is observed that

SG1 pressure is well reproduced by both codes but SG2 pressure is better reproduced

by TRACE than RELAP5 code.

• The analysis of heat transfer along the transient shows an interesting behavior, Figure

10 and Table 7. It is observed that both SGs have the same heat transfer capability

in the first stage of the test. However, when pressure reaches 2 bar in SG1 and goes

on growing in SG2, the heat transfer changes in both SG: in SG2 the heat transfer

decreases from 37% to 5% of core power, whereas in SG1 increases from 37% to 77%.

It also important to remark that both codes show that primary heat losses are quite

important in this kind of transient in experimental facilities, near 15% at high temperature in PKL facility. In real nuclear power plants, the primary heat losses are lower

because the relationship between surface and volume is lower than in experimental

facilities.

4-3
In general, test E3.1 is well reproduced. However, mass distribution problems are observed,

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

Figure 6: Pressurizer pressure. Test E3.1

Figure 7: Core exit liquid temperature. Test E3.1

4-4
Figure 8: Steam generator 1 vapor mass flow. Test E3.1

Figure 9: Steam generators pressure. Test E3.1

4-5
Figure 10: Heat transfer to steam generators and heat losses in primary side. Test E3.1.

TRACE code

Component TRACE RELAP5

10000 s 30000 s 10000 s 30000 s

SG 1 (%) 37.0 75.0-77.0 33.9 77.0

SG 2 (%) 37.0 5.0 33.9 4.0

SG 3 (%) 2.0 2.0 2.0 2.0

SG 4 (%) 2.0 2.0 2.0 2.0

Primary heat losses (%) 10.0 14.0 13 15

Total heat transfer (%) 88.0 98.0-100.0 85.0 100

Total core power (%) 100.0 100.0 100.0 100.0

Table 7: Equilibrium heat transfer balances in the primary side before and after transient

phase two for both codes. PKL E3.1

4-6
Figure 11: Pressurizer water level. Test E3.1

Figure 12: Vessel collapsed water level. Test E3.1

4-7
Figure 13: U-tubes water level, steam generator 1. Test E-3.1

Figure 14: U-tubes water level, steam generator 2. Test E-3.1

4-8
5 SIMULATION OF LOSS OF RHRS SEQUENCES

AT MIDLOOP CONDITIONS FOR 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 15, 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 guide tubes, Figure 16.

• 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 E3.1 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: 1 steam generator full of water and 2 full of air (1SG).

• Secondary pressure: 1 bar

• No Auxiliary Feedwater available.

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

PKL experimental data, Figures 17 to 20. During the transient the pressure equilibrium

condition is reached after 10000 s, lower than in PKL E3.1 (T= 15000 s). In this case, it

must be taken into account that in PKL test E3.1 there is a second steam generator full

of water which is pressurized along the transient. This aspect may explain the difference

between both times.

With respect to the equilibrium pressure, it is observed that the value of Almaraz NPP case

is higher than PKL data, Table 8. This discrepancy could be explained because during the

adjustment of PKL model it was observed that small differences in the hot/cold leg heights

and in heat losses values gives values between 4 bar and 9 bar for equilibrium primary

pressure. So, it could be expected that the differences between PKL and Almaraz NPP

gives different equilibrium values. In this sense, there is also a reference of Y. Ferng and

S. Ma (Investigation of system responses of the Maanshan nuclear power plant to the loss

of residual heat removal during midloop operations using a RELAP5/MOD3 simulation.

Nuclear Technology 116, 160-172. 1996.) which shows similar values than Almaraz NPP

model for similar sequences.

5-2
Case Maximum Pressure (bar) Time to 2 bar (min.) Time to 3 bar (min.)

1/3 SG Almaraz 11.5 33 48

1/4 SG PKL 4 – –

2/4 SG PKL – 83 165

Table 8: Simulations results for the loss of RHRS at midloop level and closed RCS. Almaraz

NPP model and experimental data from PKL test E3.1

j604

PIPE

605

TEE

604

j605 j606

FILL

FILL

j802 j801

j800

PIPE 803

802

800

801

TEE

j803

FILL

FILL

j801 j802

j800

803 PIPE

802

800

801

TEE

j803

TEE 206

PUMP 240

j207

j222

j210

j211

j215 j213 j212

FILL

270

PIPE

221

PIPE

125

222

PIPE PIPE

224

TEE 251 TEE 250

FILL

261

PIPE 260

j214

j216

j219

j220

VALVE PIPE 231 262

j217

j218

TEE 253 TEE 252

280

FILL

j304

j305

PIPE 106

PUMP 140

j107

j122

j125

j110

j111

j113

j112

j114 j115 j117

TEE 135

FILL

170

FILL

192

PIPE 131

PIPE

121

PIPE

125

122

PIPE PIPE

124

TEE 151 TEE 152

FILL

161

PIPE 160

VALVE 162

j116

j118

j119

j120

TEE 153 j104

j105

j204

VESSEL 1

j121

TEE

606

j607

j606 j609

TEE

607

TEE

608

TEE

706

j707

j708

TEE

707

TEE

708

PIPE 306

j351 j313 j315

FILL

370

TEE 351 TEE 352 PIPE 353

FILL

361

PIPE 360

VALVE 362

j314

j316

j317

j318

PUMP 340

j307

j321

j322

j325

j310

j311

PIPE

321

PIPE

325

322

PIPE PIPE

324

PIPE 331

TEE

806

j807

j806 j809

TEE

807

TEE

808

j806

j804

PIPE

805

TEE

804

j805

j704

PIPE

705

TEE

704

j705

j221 j125

j709

j706 PIPE

809

j810

709

j710

PIPE

j603

FILL

FILL

j602 j601

j600

PIPE 603

602

600

601

TEE

PIPE 401

PIPE 402

TEE 414

PIPE

400

j400

j401

j403

j402

j404

j415

j205

SPRAY LINES

SAFETY AND RELIEF VALVES

STEAM LINE 3

j610

609

PIPE

STEAM LINE 1

STEAM LINE 1

U−TUBES

U−TUBES

U−TUBES

Figure 15: Nodalization of the TRACE model of Almaraz NPP

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 16: Nodalization of the TRACE model of 3D VESSEL

5-4
0 2000 4000 6000 8000 10000

Time (s)

2e+05

4e+05

6e+05

8e+05

1e+06

1,2e+06

1,4e+06

pressure (Pa)

Primary Pressure. 1 SG

RWST gravity feed (2bar)

RWST gravity feed (3bar)

Figure 17: RCS pressure. Transients similar to PKL E3.1

2000 4000 6000 8000

Time (s)

0

2

4

6

8

10

level (m)

1SG case. Ascendent U-tubes.

1SG case. Descendent U-tubes.

Figure 18: U-tubes level of SG operating. Transients similar to PKL E3.1

5-5
2000 4000 6000 8000

Time (s)

6

7

8

Level (m)

Vessel level. 1SG

Active core level

Figure 19: Vessel collapsed water level. Transients similar to PKL E3.1

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.

0 5000 10000 15000 20000 25000 30000 35000

Simulation time (s)

0

50000

1e+05

1,5e+05

2e+05

2,5e+05

3e+05

CPU time (s)

Figure 20: Execution time. PKL test E3.1. TRACE model

6-1

7 CONCLUSIONS

The PKL III test facility replicates the entire primary system and most of the secondary

system of a typical 1300 MWe pressurized water reactor of Siemens / KWU design. In

test E3.1, the primary inventory was at midloop level (∼ 58%) and thus, the influence of

noncondensables on two-phase cooling was investigated. In this test was also analyzed the

capability of the reflux-condensation as a mechanism of cooling in a closed RCS considering

the availability of two steam generators, one of them operable.

In this report, a post-test analysis of PKL III E3.1 using RELAP5/MOD3 and TRACE

codes has been presented. The main findings of the comparison of RELAP5/Mod 3.2 and

TRACE 5.0 results with the PKL III E3.1 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.

• All participants of the E3.1 benchmark obtained a wrong primary mass distribution,

with a very large water level in the PZR. Probably, 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).

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

results that PKL test have been obtained.

7-1