NUREG/IA-0426 

International 

 Agreement Report

Simulation of LSTF Upper Head Break 

(OECD/NEA ROSA Test 6.1) with TRACE Code. 

Application to a PWR NPP Model 

Prepared by:

César Queral1, Juan Gonzalez-Cadelo1, Gonzalo Jimenez1, 

Ernesto Villalba1

, Julio Perez2

 

 1Universidad Politécnica de Madrid

Alenza 4. 28003 Madrid.

Spain.

2

Consejo de Seguridad Nuclear

Pedro Justo Dorado Dellmans, 11, 28040 Madrid.

Spain.

 

Kirk Tien, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: December 2013

Date Published: March 2014

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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infringe privately owned rights. 
NUREG/IA-0426 

International 

 Agreement Report

Simulation of LSTF Upper Head Break 

(OECD/NEA ROSA Test 6.1) with TRACE Code. 

Application to a PWR NPP Model 

Prepared by:

César Queral1, Juan Gonzalez-Cadelo1, Gonzalo Jimenez1, 

Ernesto Villalba1

, Julio Perez2

 

 1Universidad Politécnica de Madrid

Alenza 4. 28003 Madrid.

Spain.

2

Consejo de Seguridad Nuclear

Pedro Justo Dorado Dellmans, 11, 28040 Madrid.

Spain.

 

Kirk Tien, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: December 2013

Date Published: March 2014

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 LSTF Upper Head Break experiment (OECD/NEA ROSA test 6.1) has been

performed with TRACE code. This test, included within the OECD/NEA ROSA project,

attempts to analyze the phenomenology and different accident management actions after

the occurrence of a Upper Head break with failure of High Pressure Safety Injection (HPSI).

The comparison between the experimental data and the results obtained with TRACE code

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

Additionally, a broad analysis of Upper Head Small Break Loss of Coolant Accident (SBLOCA)

with HPSI failed in a Westinghouse PWR has been performed taking into account different

accident management actions and conditions in order to check their suitability.

These works has been performed in the framework of OECD/NEA ROSA and CAMP

projects.

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
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.

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 OECD/NEA PKL 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 OECD/NEA ROSA 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 OECD/NEA PKL and OECD/NEA ROSA 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

CONTENTS . . . . . . . . . . . . . . . . . . . . . . vii

FIGURES . . . . . . . . . . . . . . . . . . . . . . . ix

TABLES . . . . . . . . . . . . . . . . . . . . . . . xiii

EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . xv

ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . xvii

ABBREVIATIONS . . . . . . . . . . . . . . . . . . . xix

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

2 DESCRIPTION OF LSTF AND OECD/NEA ROSA TEST 6.1 . . . 2-1

2.1 Description of LSTF . . . . . . . . . . . . . . . . . 2-1

2.2 Description of OECD/NEA ROSA Test 6.1 . . . . . . . . . . 2-8

3 DESCRIPTION OF TRACE MODEL OF LSTF . . . . . . . . 3-1

3.1 First Stage: LSTF model translation from TRAC-PF1 to TRACE . . . 3-2

3.2 Second Stage: LSTF TRACE model improvements . . . . . . . . 3-9

4 COMPARISON OF EXPERIMENTAL RESULTS AND SIMULATIONS

OF OECD/NEA ROSA TEST 6.1 . . . . . . . . . . . . . 4-1

4.1 Sensitivity analysis to RCP trip delay . . . . . . . . . . . . 4-11

5 SIMULATION OF UPPER HEAD SBLOCA IN ALMARAZ NPP . 5-1

5.1 Description of Almaraz NPP model . . . . . . . . . . . . . 5-4

5.2 Comparison between Almaraz NPP and LSTF . . . . . . . . . 5-7

5.3 Comparison between Almaraz NPP model and LSTF experimental results . 5-10

5.4 Sensitivity analysis in Almaraz NPP model . . . . . . . . . . 5-14

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

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

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

viii
FIGURES

Page

Figure 1 Large Scale Test Facility (LSTF) . . . . . . . . . . . . . 2-2

Figure 2 LSTF primary side (plan view) . . . . . . . . . . . . . 2-3

Figure 3 LSTF pressure vessel (elevation view) . . . . . . . . . . . 2-4

Figure 4 LSTF steam generator (elevation view) . . . . . . . . . . . 2-5

Figure 5 LSTF pressurizer (elevation view) . . . . . . . . . . . . 2-6

Figure 6 LSTF hot and cold legs (elevation view) . . . . . . . . . . 2-7

Figure 7 Evolution of main variables in OECD/NEA ROSA Test 6.1 . . . . 2-10

Figure 8 TRAC-PF1/MOD1 model of LSTF (JAERI) . . . . . . . . . 3-1

Figure 9 TRACE model of LSTF Vessel . . . . . . . . . . . . . 3-3

Figure 10 TRACE model of LSTF Steam Generator . . . . . . . . . . 3-4

Figure 11 TRACE model of LSTF Hot and Cold Legs . . . . . . . . . 3-5

Figure 12 TRACE model of LSTF Pressurizer . . . . . . . . . . . . 3-6

Figure 13 TRACE model of LSTF ECCS . . . . . . . . . . . . . 3-7

Figure 14 TRACE model of LSTF Steam Lines . . . . . . . . . . . 3-8

Figure 15 TRACE 2D model of LSTF pressurizer . . . . . . . . . . . 3-12

Figure 16 Liquid temperatures: 1D and 2D pressurizer models . . . . . . . 3-13

Figure 17 LSTF Vessel - Improved nodalization with 19 levels . . . . . . . 3-14

Figure 18 LSTF Hot and cold legs, and vessel bypass of improved model . . . . 3-15

Figure 19 Improved LSTF steam generator model . . . . . . . . . . . 3-16

Figure 20 TRACE model of LSTF steam generator with 9 U-tubes . . . . . 3-17

Figure 21 TRACE Model of LSTF break rupture discharge valve . . . . . . 3-17

Figure 22 Comparison between volumes of improved model and LSTF PZR . . . 3-18

Figure 23 Comparison between volumes of improved model and LSTF vessel . . 3-19

Figure 24 Comparison between volumes of improved model and LSTF legs . . . 3-20

Figure 25 Comparison between volumes of improved model and LSTF SG . . . 3-21

Figure 26 TRACE model of LSTF . . . . . . . . . . . . . . . 3-23

Figure 27 SNAP mask of TRACE model of LSTF . . . . . . . . . . 3-24

Figure 28 Primary pressure. Comparison between experimental and 1/9 U-tubes models 4-3

Figure 29 PCT. Comparison between experimental and 1/9 U-tubes models . . 4-4

Figure 30 CET. Comparison between experimental and 1/9 U-tubes models . . 4-4

ix
Figure 31 Loop B mass flow. Comparison between experimental and 1/9 U-tubes models 4-5

Figure 32 Break integrated mass flow. Experimental and 1/9 U-tubes models . . 4-5

Figure 33 PCT. Experimental data and models with adjusted RFRIC and reflood option 4-6

Figure 34 PCT for test 6.1 (MODEL-4) . . . . . . . . . . . . . . 4-6

Figure 35 Primary and secondary pressures for test 6.1 (MODEL-4) . . . . . 4-7

Figure 36 Core liquid level for test 6.1 (MODEL-4) . . . . . . . . . . 4-7

Figure 37 Break mass flow for test 6.1 (MODEL-4) . . . . . . . . . . 4-8

Figure 38 Temperature at core exit for test 6.1 (MODEL-4) . . . . . . . 4-8

Figure 39 Void fractions. Snapshots of SNAP video simulation (1/2) . . . . . 4-9

Figure 40 Void fractions. Snapshots of SNAP video simulation (2/2) . . . . . 4-10

Figure 41 Simplified scheme of standar Westinghouse EOPs in LOCA scenaries . 4-11

Figure 42 Core power with and without core protection system . . . . . . 4-12

Figure 43 Pump rotational speed for different RCP trip times . . . . . . . 4-12

Figure 44 Integrated break mass flow for different RCP trip times . . . . . . 4-13

Figure 45 Core level for different RCP trip times . . . . . . . . . . . 4-13

Figure 46 PCT for different RCP trip times . . . . . . . . . . . . 4-14

Figure 47 PCT, with and without core protection system . . . . . . . . 4-14

Figure 48 Scheme of Almaraz NPP (plan view) . . . . . . . . . . . 5-1

Figure 49 Scheme of Almaraz NPP (elevation view) . . . . . . . . . . 5-2

Figure 50 TRACE model of Almaraz NPP (SNAP mask with mean components) . 5-5

Figure 51 Comparison between Almaraz NPP and LSTF VESSEL components . . 5-9

Figure 52 Integrated break flow. Almaraz NPP vs 1/48 scaled LSTF . . . . . 5-10

Figure 53 Primary and secondary pressure. Almaraz NPP vs 1/48 scaled LSTF . 5-11

Figure 54 Core power comparison. Almaraz NPP vs 1/34.52 scaled LSTF . . . 5-11

Figure 55 Integrated break flow. Almaraz NPP vs 1/34.52 scaled LSTF . . . . 5-12

Figure 56 Primary and secondary pressure. Almaraz NPP vs 1/34.52 scaled LSTF 5-12

Figure 57 Peak cladding temperature. Almaraz NPP vs 1/34.52 scaled LSTF . . 5-13

Figure 58 Integrated break flow. Almaraz NPP vs LSTF with different DC . . . 5-16

Figure 59 PCT. Almaraz NPP vs LSTF with different DC . . . . . . . . 5-16

Figure 60 Integrated break flow. Almaraz NPP vs LSTF with different break areas 5-17

Figure 61 PCT sensitivity to break area . . . . . . . . . . . . . . 5-17

Figure 62 PCT sensitivity to break location in different radial sectors . . . . 5-18

x
Figure 63 Integrated flow from ACCs. Almaraz NPP vs LSTF. Break in sectors 3/4 5-18

Figure 64 RCS and SG pressures. Almaraz NPP vs LSTF with break in sector 4 . 5-19

Figure 65 Integrated flow from ACCs. Almaraz NPP vs LSTF. Different friction factors 5-19

Figure 66 PCT sensitivity to friction factors at ACC outlet . . . . . . . . 5-20

Figure 67 PCT sensitivity to friction Upper Downcomer Area . . . . . . . 5-20

Figure 68 Primary and secondary pressures. Sensitivity to friction factors at guide tubes 5-21

Figure 69 CPU time performance for ROSA/LSTF test 6.1 simulation . . . . 6-1

Figure 70 CPU time performance for Almaraz NPP model simulation of test 6.1 . 6-2

xi

TABLES

Page

Table 1 Initial conditions in OECD/NEA ROSA Test 6.1 . . . . . . . . 2-9

Table 2 Chronology of major events in OECD/NEA ROSA Test 6.1 . . . . . 2-9

Table 3 Heat losses. Comparison between LSTF and TRACE model . . . . 3-10

Table 4 Added U-tubes geometry in steam generator model . . . . . . . 3-11

Table 5 Comparison between volumes of JAEA data and ROSA improved model . 3-22

Table 6 Initial and boundary conditions for OECD/NEA ROSA test 6.1 . . . 4-2

Table 7 Chronology of events: OECD/NEA ROSA Test 6.1 and TRACE simulation 4-3

Table 8 Main operating parameters of the Almaraz I NPP . . . . . . . . 5-3

Table 9 Comparison among LSTF, Almaraz NPP and Tsuruga NPP (1/2) . . . 5-7

Table 10 Comparison among LSTF, Almaraz NPP and Tsuruga NPP (2/2) . . . 5-8

Table 11 CPU performance for LSTF and Almaraz NPP simulations of test 6.1 . 6-1

xiii

EXECUTIVE SUMMARY

The Large Scale Test Facility (LSTF) is a full-height, full-pressure and 1/48 volumetrically scaled simulator for a Westinghouse-type 4-loop (3423 MWt) pressurized water reactor

(PWR) with primary and secondary coolant systems including an electrically-heated simulated core, emergency core cooling system (ECCS) and control systems for accident management (AM) actions. The maximum core power of 10 MWt is equivalent to 14% of the

1/48-scaled PWR rated power covering the scaled PWR decay heat after the scram.

OECD/NEA ROSA test 6.1 (SB-PV-09) was carried out on November 17, 2005 in the Large

Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). This test consisted of a vessel upper-head break equivalent in size to a 1.9% cold leg break, and without

availability of High Pressure Safety Injection.

In this report, an extended analysis of Upper Head SBLOCA with HPSI unavailability in a

Westinghouse PWR is presented. The analysis has been performed through 2 stages:

1. In a first stage, the post-test simulation with TRACE code was performed, and then

extensively evaluated through comparision with the experimental results.

2. In a second stage, similar transients to test 6.1 have been simulated with the TRACE

model of Almaraz NPP (Westinghouse 3 loop design). This extended analysis takes

into account different accident management actions and conditions in order to check

their suitability.

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

to properly simulate transient conditions. The main findings of the comparison of TRACE

results with the OECD/NEA ROSA test 6.1 experiment are:

• Results of OECD/NEA ROSA Test 6.1 have been well reproduced through successive

improvements and modifications in the LSTF model. Main changes in order to obtain

an adequate post-test simulation were RFRIC factors and the activation of reflood

model, as well as a renodalization of the vessel, cold and hot legs and steam generators.

• Not all improvements added to the LSTF model have lead to good results, as expected.

For example, addition of 9 U-tubes in the SG of LSTF model yields worse results than

the simplest model, with only 1 U-tube.

• One of the pre-established experimental conditions of OECD/NEA ROSA Test 6.1,

was the trip of reactor coolant trip at the same time that break occurs. Sensitivity

analysis to RCP trip delay shows that maximum cladding temperature increases as

RCP trip delay increases, but only until approximately 1000 seconds after the break.

xv
Beyond that point, PCT decreases slightly, remaining almost constant independently

of RCP trip delay.

• Transposition of Test 6.1 into Almaraz NPP requires taking into account respective

scaling factors. Once considered this scale factor, both transients look very similar,

except some deviations in pressure and PCT. Depressurization is quite similar in both

models, until ACCs demand, because of an earlier discharge in Almaraz due to its higher

pressure set-point. Cladding temperature begins to rise up earlier in experimental test

PCT is significantly lower in Almaraz simulations, due to its earlier ACCs discharge.

• Simulation of Test 6.1 in Almaraz NPP model shows diverse grades of sensitivity to several parameter modifications. Break location, steady state upper head mass flow and

friction factors at ACCs exit lead to little change in results. Model exhibits medium

sensitivity to discharge coefficients and upper downcomer area, and shows high sensitivity to break area size, RCP trip delay and quantity of ACCs available.

xvi
ACKNOWLEDGMENTS

This work contains findings that were produced within the OECD/NEA ROSA Project. The

authors are grateful to the Management Board of the ROSA Project for their consent to this

publication, and thank Almaraz-Trillo AIE and the Spanish Nuclear Regulatory Body (CSN)

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

xvii

ABBREVIATIONS

ACC Accumulator

AFWS Auxiliary Feedwater System

AM Accident Management

CAMP Code Applications and Maintenance Program

CET Core Exit Thermocouples

CRDM Control Rod Drive Mechanism

CSN Consejo de Seguridad Nuclear (Spanish Nuclear Council)

CSNI Committee on the Safety of Nuclear Installations

CVCS Chemical and Volume Control System

DC Discharge Coefficient

ECCS Emergency Core Cooling System

EOP Emergency Operating Procedure

FWS Feedwater System

HPSI High Pressure Safety Injection

JAEA Japan Atomic Energy Agency

JAERI Japan Atomic Energy Research Institute

LBLOCA Large Break Loss of Coolant Accident

LOCA Loss Of Coolant Accident

LOFT Loss of Fluid Tests

LPSI Low Pressure Safety Injection

LSTF Large Scale Test Facility

NRC Nuclear Regulatory Commission

NPP Nuclear Power Plant

NSSS Nuclear Steam Supply System

OECD Organisation for Economic Co-operation and Development

PCT Peak Cladding Temperature

PKL Prim ̈ar-Kreis-Lauf

PORV Pilot Operated Relief Valve

PWR Pressurized Water Reactor

PZR Pressurizer

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RHRS Residual Heat Removal System

RPV Reactor Pressure Vessel

ROSA Rig Of Safety Assessment

RV relief Valve

SBLOCA Small Break LOCA

SESAR Senior Group of Experts on Nuclear Safety Research

SG Steam Generator

SI Safety Injection

SNAP Symbolic Nuclear Analisys Package

xix
TH Thermal-Hydraulics

TMI-2 Three Mile Island NPP - Unit 2

UH Upper Head

UPM Universidad Politecnica de Madrid

USNRC United States Nuclear Regulatory Commission

W-4L Generic Westinghouse 4-loop PWR

xx
1 INTRODUCTION

In the beginning of PWR transient analysis history, the most limiting accident considered

was the surge line break (10 inch), because credit was not given to an RCS large pipe break.

In 1966, the Atomic Energy Commission required the break analysis of the larger pipe in

the NSSS: the RCS piping double-end break. For many years, it was thought to be the most

conservative accident for PWRs.

In 1979, the Small Break Loss of Coolant Accident (SBLOCA) at Three Mile Island 2 NPP,

shown that SBLOCA had to be taking into account as the plant behavior is quite different

from LBLOCA and it is not fully covered by the classical LBLOCA analysis.

The most limiting SBLOCA since that time is the break at the cold leg, as it drives to

the complete loss of one coolant injection path. Later on, the finding of a vessel head wall

thinning at the Davis Besse reactor in 2002 showed the possibility of a SBLOCA in the upper

head of the reactor vessel due to the circumferential cracking of a CRDM penetration nozzle.

With the aim of simulating PWR behavior during such type of scenarios, several experimental

test has been performed in LSTF facility. The Large Scale Test Facility (LSTF) is a fullheight, full-pressure and 1/48 volumetrically scaled simulator for a Westinghouse-type 4-loop

PWR of 3423 MWt with primary and secondary coolant systems including an electricallyheated simulated core, ECCS and control systems for accident management actions. The

maximum core power of 10 MW is equivalent to 14% of the 1/48-scaled PWR rated power

covering the scaled PWR decay heat after the scram.

The OECD/NEA ROSA Project, which started in 2005 by the agreement between JAEA,

OECD/NEA and thirteen member countries, determined to conduct a SBLOCA test (Test

6.1, SB-PV-09 in JAEA). This test 6.1 simulates a PWR vessel upper-head break equivalent

in size to a 1.9% cold leg break. The objective of the test was to study the effect of accident

management (AM) actions and to provide integral test data for assessment and development

of advanced analytical codes.

In this report, a post-test analysis of OECD/NEA ROSA test 6.1 using TRACE 5.0 code

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 Westinghouse 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 upper head small break LOCA transients.

1-1

2 DESCRIPTION OF LSTF AND OECD/NEA

ROSA TEST 6.1

2.1 Description of LSTF

LSTF is a large scaled-down model of a four loop Westinghouse PWR design of 3423 MWt

(Reference plant: Tsuruga NPP Unit 2), Figures 1 to 6. Its main characteristics are:

• LSTF has 2 loops, instead the 4 loops of the reference PWR.

• Elevations are scaled 1/1 while the volumes, power and mass flows are scaled 1/48 for

the Reactor Coolant System (RCS) and 1/145 for the secondary system.

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

power of 10 MWt (14% of Tsuruga-2 rated power).

• Maximum operating pressure is 180 bar.

• Each steam generator has 141 U-tubes (4.2% of Tsuruga-2 U-tubes for SG), of original

size and material.

As LSTF is full scale in height, the natural circulation phenomena during accident conditions

is suitable to reproduce, so most of the different behaviors of a PWR during accident can be

estimated with fidelity.

One of the most interesting studies done usually in the facility are those related with the

role of operators in the accident management. As the control systems are implemented, the

strategies are quite much the same as in a real plant transient, so the impact of the different

timing and implementation of the manual actions can be measured on line as the plant status

is monitored during the transients.

Both, the ability to reproduce the thermal-hydraulic phenomena and the operator actions

allow the detailed analysis of several accident scenarios like small break LOCA, steam generator U-tube rupture,main steam line break, etc.

2-1
Figure 1: Large Scale Test Facility (LSTF)

2-2
Figure 2: LSTF primary side (plan view)

2-3
Figure 3: LSTF pressure vessel (elevation view)

2-4
Figure 4: LSTF steam generator (elevation view)

2-5
Figure 5: LSTF pressurizer (elevation view)

2-6
Figure 6: LSTF hot and cold legs (elevation view)

2-7
2.2 Description of OECD/NEA ROSA Test 6.1

The main objective of OECD/NEA ROSA Test 6.1 is to analyze the phenomenology of

upper head SBLOCA with HPSI failed in a PWR reactor. This test is also useful to obtain

experimental thermal-hydraulic data for the assessment of thermal-hydraulic computer codes

like TRACE.

Other conditions and specifications for the test are the following:

1. Total failure of HPSI system when demanded.

2. Loss of off-site power coinciding with scram.

3. AFW system initiated immediately after SI signal, at primary pressure of 12.27 MPa

(when required AM action to maintain SG secondary-side water level about U-tubes).

4. Initiation of steam generator (SG) secondary-side depressurization by fully opening

relief valves, as accident management (AM) action when core exit temperature reaches

623 K.

5. Break size is 1.9% cold leg equivalent (diameter of 13.8 mm).

6. The core power is automatically decreased by the core protection system when the

maximum fuel rod surface temperature exceeds a certain maximum. Threshold temperature for the LSTF core protection system set as follows: 958 K = 75%, 968 K =

50%, 969 K = 25%, 970 K = 10% and 973 K = 0% of pre-determined value.

Steady-state conditions of the test can be seen in Table 1. Chronology of major events in

Test 6.1 is shown in Table 2 and Figure 7.

The relatively big size of the break results at the beginning of the transient, in a fast depressurization and loss of inventory in the RCS, leading to an early core uncovery. Subsequent

core overheating triggers the AM action of depressurizing the secondary side when the Core

Exit Thermocouples (CETs) detect a high temperature (T>623K). However, the fact that

primary pressure was much lower than the SG secondary-side pressure (see Figure 7) indicates that detection of high CET was late in time, and so subsquent AM action of opening

SG reielf valves.

This AM action was ineffective in the early stage,as the LSTF core protection system automatically had to decrease the core power down to 10% of the decay power level as the

maximum fuel rod surface temperature exceeded the core protection limit (T>958 K).

2-8
The test had to be terminated prematurely to avoid excessive overheating of the core. Results

showed that the core uncovery had started significantly early before the CET thermocouples

indicated superheating and that the temperature increase rate was higher in the core than

in the CET. The results suggested that the response of the CET thermocouples could be

inadequate to initiate the relevant AM actions.

PARAMETER VALUE

Power (MW) 10.0 ± 0.07

Hot Leg temperature (K) 598.1 ± 2.75

Cold Leg temperature (K) 562.4 ± 2.75

PZR pressure (MPa) 15.5 ± 0.108

PZR level (m) 7.2 ± 0.25

Mass flow per loop (kg/s) 24.3 ± 1.25

Secondary pressure (MPa) 7.3 ± 0.054

Steam mass flow (kg/s) 2.74 ± 0.10

Feed water mass flow (kg/s) 2.74 ± 0.05

ACCs pressure (MPa) 4.51 ± 0.054

LPSI pressure (MPa) 1.24 ± 0.108

Table 1: Initial conditions in OECD/NEA ROSA Test 6.1

TIME (s) EVENT

2000 Break valve open

2022 SCRAM signal (Primary Pressure = 12.97 MPa)

2028 SI signal (Primary pressure = 12.27 MPa)

2050 Break flow from single-phase liquid to two-phase flow

2276 Primary coolant pumps stop

2700 Break flow to single-phase vapor

∼ 2800 Primary pressure lower than SG secondary-side pressure, Core uncovery

3090 Initiation of SG secondary-side depressurization (full opening of relief

valves, core exit temperature = 623 K)

∼ 3200 Core power decrease by LSTF core protection system (Max. fuel rod

surface temperature = 970 K)

∼ 3300 Initiation of accumulator system (Primary pressure = 4.51 MPa)

∼ 4300 Inflow of nitrogen gas from accumulator tank into primary loop

∼ 4900 Initiation of LPSI system (RPV lower plenum pressure = 1.23 MPa)

5265 Break valve closure

Table 2: Chronology of major events in OECD/NEA ROSA Test 6.1

2-9
Figure 7: Evolution of main variables in OECD/NEA ROSA Test 6.1

2-10
3 DESCRIPTION OF TRACE MODEL OF LSTF

The group of Universidad Politecnica de Madrid (UPM) is working with TRACE model of

LSTF facility since February 2006. The development of the TRACE model of LSTF is based

on the TRAC-PF1 model (see Figure 8) delivered by the Japan Atomic Energy Research

Institute (JAERI) to the participants of OECD/NEA ROSA project.

Figure 8: TRAC-PF1/MOD1 model of LSTF (JAERI)

Nowadays, TRACE model of LSTF contains following systems and equipments:

• Primary system:

– Vessel

– Steam generators U-tubes (primary-side)

– Hot legs, intermediate legs and cold legs

– Pressurizer and surge line

– Reactor coolant pumps

• Secondary system:

– Steam generators (secondary-side)

3-1
– Main steam lines

• Safeguard systems:

– Accumulators system

– High and low pressure safety injection systems

– Auxiliary feedwater system

• Instrumentation and control systems:

– Primary: Reactor trip logic; Reactor coolant pump trip control; Pressurizer level

and pressure control; High and low pressure safety injection fail control and Accumulators signal control.

– Secondary: Turbine trip logic; Secondary-side pressure and level control; Relief

valves control (for secondary depressurization purposes); Main feedwater system

activation control and Auxiliary feedwater system activation control.

Initially, in a first stage, LSTF TRAC-PF1 model had to be adapted to TRACE code. Later

on, in a second stage, since 2006 many improvements have been included. This section

contains a briefly description of both phases.

3.1 First Stage: LSTF model translation from TRAC-PF1 to TRACE

LSTF model for TRAC-PF1 code (Figure 8) was migrated to TRACE code.Main tasks

related with the translation of the model were:

1. Old VESSEL component was translated to TRACE: VESSEL component from TRACPF1 model was divided into hydraulic and thermal components, creating new HTSTR

to substitute the core heaters included in component VESSEL (see Figure 9).

2. Old STGEN component was translated to TRACE model as a set of components (TEEs

and PIPEs), preserving volumes and lengths (see Figure 10).

3. The steam generators recirculation ratio was adjusted.

4. Total mass flow was adjusted in primary loops using FRIC parameters and rated head

in RCP.

5. Volume vs. height plots were checked with respect the facility data and some discrepancies were corrected.

3-2
Figure 9: TRACE model of LSTF Vessel

3-3
Figure 10: TRACE model of LSTF Steam Generator

3-4
Figure 11: TRACE model of LSTF Hot and Cold Legs

3-5
Figure 12: TRACE model of LSTF Pressurizer

3-6
Figure 13: TRACE model of LSTF ECCS

3-7
Figure 14: TRACE model of LSTF Steam Lines

3-8
3.2 Second Stage: LSTF TRACE model improvements

Later on, in a second stage, LSTF TRACE model was upgraded with the addition of several

improvements to the model:

1. New 2D pressurizer model (see Figure 15) was included in order to simulate adequately

the natural circulation inside the pressurizer and also to avoid excessive cooling in upper

cells of the model during long quasi steady state transients (see Figure 16).

2. Pressurizer level and pressure control systems were added to set the steady state more

adequately.

3. New detailed proportional and base heaters with more detail were also added in the

pressurizer.

4. Heat losses of the whole model were adjusted (see Table 3).

5. The mass flow rate from the downcomer to the upper head of the vessel was adjusted

to the specified one (0.3% of the downcomer vessel total mass flow).

6. The temperature in the Upper Head of the Vessel was adjusted to the measured one

(about 586 K).

7. ΔP along the model was revised with reasonable results.

8. Control blocks, signal variables and trips were renumbered to avoid misunderstandings

reading output data.

9. New signal variables to measure heat losses, surface temperature in core heaters, liquid

level in upper head, core and PZR.

10. Several masses of HTSTR components were corrected, i.e. U-tube support plate.

11. The OFFTAKE model was activated in the connections of the valves that simulate

breaks in different localizations of LSTF model.

12. Several models of break rupture discharge valve were tested for simulation of test 6.1.

13. New SG 9 heights U-tube model, with a more detailed nodalization (see Figure 20, and

Table 4).

14. Adjustment of several RFRIC factors.

15. Reflood model was activated.

16. Improved nodalization and model dimensions for better correlation between model

height/volume and LSTF height/volume (see Table 5):

3-9
• Modified vessel with 19 levels, improved nodalization and more adjusted lower

and upper plenum, core, lower and upper head and downcomer (see Figure 17).

• Modified cold and hot legs in loop A and loop B (See Figure 18).

• Improved steam generators modelling (See Figure 19).

17. Finally, an animation mask was created with SNAP tool (see Figure 27). This mask

allows performing videos of the simulations, which allows an easy interpretation of the

transient behavior.

Modifications 1 to 12 lead to a first model (MODEL-1). Later, a second model with nine

U-tube elevations was obtained in order to check the impact of several heights in the results

(MODEL-2). As it will be shown later, this model did not improve results of MODEL-1.

Two new modificactins (14 and 15), included in MODEL-3, improved the results although

maintining slight deviations. Finally, a full review of nodalization and model dimensions was

performed obtaining the last model (MODEL-4) that provided the best results. Nowadays,

the TRACE model of ROSA/LSTF facility (see Figure 26) has 178 thermal-hydraulic components (2 VESSEL, 45 PIPE, 8 TEE, 2 SEPD, 22 VALVE, 2 PUMP, 9 FILL, 15 BREAK,

70 HTSTR and 3 POWER), 1013 Signal Variables, 167 Control Blocks and 20 Trips.

Component Temp(K) LSTF (kW) TRACE (kW)

Pressurizer 620 15.0 15.41

Pressure Vessel 600 58.6 59.5

Primary Loop A 600 19.0 19.5

Primary Loop B 600 19.0 19.5

SG-A Plena 600 2.6 2.96

SG-B Plena 600 3.2 2.96

SG-A Sec. Vessel 560 24.1 28.05

SG-B Sec. Vessel 560 31.1 28.05

SG-A Downcomer Piping 560 3.5 3.27

SG-B Downcomer Piping 560 2.2 3.27

Table 3: Heat losses of the ROSA/LSTF System. Comparison between LSTF and TRACE

model

3-10
Type R (mm) L (mm) Number

1 50,8 9439,9 21

2 83,3 9590,7 19

3 115,8 9741,2 19

4 148,3 9891,7 19

5 180,8 10042,2 17

6 213,3 10192,7 15

7 245,8 10343,2 13

8 278,3 10493,7 11

9 310,8 10644,2 7

total 141

Table 4: Added U-tubes geometry in steam generator model

3-11
Figure 15: TRACE 2D model of LSTF pressurizer

3-12
1000 1500 2000 2500

Time (s)

613

614

615

616

617

618

619

liquid temperature (K)

cell 1

cell 2

cell 3

cell 4

cell 5

T Sat

1000 1500 2000

Time (s)

617

618

619

620

saturation temperature (K)

cell 1 int

cell 1 ext

cell 2 int

cell 2 ext

cell 3 int

cell 3 ext

cell 4 int

cell 4 ext

cell 5 int

cell 5 ext

T sat

Figure 16: Liquid temperatures: 1D and 2D pressurizer models

3-13
Figure 17: LSTF Vessel - Improved nodalization with 19 levels

3-14
Figure 18: LSTF Hot and cold legs, and vessel bypass of improved model

3-15
Figure 19: Improved LSTF steam generator model

3-16
Figure 20: TRACE model of LSTF steam generator with 9 U-tubes

Figure 21: TRACE Model of LSTF break rupture discharge valve

3-17
PRESSURIZER

0

2

4

6

8

10

12

0 0.2 0.4 0.6 0.8 1 1.2 1.4

V (m3)

H (m)

TRACE

JAEA

SURGE LINE

0

1

2

3

4

5

6

7

8

9

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04

V (m3)

H (m)

TRACE

JAEA

Figure 22: Comparison between volumes of PZR TRACE improved model and PZR of LSTF

3-18
DOWNCOMER

0

1

2

3

4

5

6

7

8

9

10

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

V (m3)

H (m)

TRACE

JAEA

LOWER PLENUM

0

0.5

1

1.5

2

2.5

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

V (m3)

H (m)

TRACE

JAEA

CORE

0

0.5

1

1.5

2

2.5

3

3.5

4

4.5

0.0000 0.0500 0.1000 0.1500 0.2000 0.2500 0.3000 0.3500 0.4000 0.4500 0.5000

V (m3)

H (m)

TRACE

JAEA

UPPER PLENUM

0

0.5

1

1.5

2

2.5

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5

V (m3)

H (m)

TRACE

JAEA

UPPER HEAD

0

0.5

1

1.5

2

2.5

0 0.1 0.2 0.3 0.4 0.5 0.6

V (m3)

H (m)

TRACE

JAEA

Figure 23: Comparison between volumes of TRACE improved model and LSTF for vessel

components

3-19
HOT LEG A

0

0.05

0.1

0.15

0.2

0.25

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14

V (m3)

H (m)

TRACE

JAEA

COLD LEG A

0

0.05

0.1

0.15

0.2

0.25

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14 0.16 0.18

V (m3)

H (m)

TRACE

JAEA

Figure 24: Comparison between volumes of TRACE improved model and LSTF for hot and

cold legs

3-20
SG-A INLET PLENUM

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5

V (m3)

H (m)

TRACE

JAEA

SG-A INLET PIPE

0

0.1

0.2

0.3

0.4

0.5

0.6

0 0.002 0.004 0.006 0.008 0.01 0.012 0.014 0.016 0.018 0.02

V (m3)

H (m)

TRACE

JAEA

SG-A OUTLET PLENUM-LSA

0

1

2

3

4

5

6

7

0 0.1 0.2 0.3 0.4 0.5 0.6

V (m3)

H (m)

TRACE

JAEA

Figure 25: Comparison between volumes of TRACE improved model and LSTF for SG

components

3-21
Component Volume Volume Error Relative

TRACE (m3) JAEA (m3) (m3) Error

Lower Plenum 0.3358 0.3364 -0.0006 -0.19%

Core 0.4598 0.4477 0.0121 2.7%

Upper Plenum 0.4714 0.4950 -0.0236 -4.77%

Upper Head 0.4963 0.4963 0.0 0.0%

Downcomer 0.9354 0.9784 -0.043 -4.4%

Total RPV 2.6987 2.7538 -0.05513 -2.0%

Prizer 1.2955 1.2955 0.0 0.0%

Surge Line 0.0365 0.0365 0.000003 0.08%

Total PZR 1.332003 1.332 0.000003 0.0002%

Hot Leg A 0.1133 0.1133 0.0 0.0%

Hot Leg B 0.1155 0.1156 -0.0001 -0.086%

Cold Leg A 0.1366 0.1531 -0.0165 -10.75%

Cold Leg B 0.1366 0.1617 -0.0250 -15.5%

LSA-PCA 0.1100 0.0921 0.0179 19.43%

LSB-PCB 0.1100 0.0926 0.0174 18.79%

Total loops 0.72208 0.7284 -0.00632 -0.86%

SGA Inlet Pipe 0.0177 0.0178 -0.0001 -0.56%

SGB Inlet Pipe 0.0177 0.0178 -0.0001 -0.56%

SGA Inlet Plenum 0.4306 0.4351 -0.0045 -1.03%

SGB Inlet Plenum 0.4306 0.4371 -0.0065 -1.49%

SGA U-tubes (IN) 0.43175 0.43346 -0.0017 -0.39%

SGB U-tubes (IN) 0.43175 0.43346 -0.0017 -0.39%

SGA U-tubes (OUT) 0.43175 0.43346 -0.0017 -0.39%

SGB U-tubes (OUT) 0.43175 0.43346 -0.0017 -0.39%

SGA Outlet Plenum-LSA 0.3416 0.3409 0.000737 0.22%

SGB Outlet Plenum-LSB 0.3416 0.3395 0.002137 0.63%

Total SGs primary 3.3069 3.3222 -0.0153 -0.46%

TOTAL PRIMARY 8.0596 8.1364 -0.0768 -0.94%

Table 5: Comparison between volumes of JAEA data and ROSA improved TRACE model

3-22
Figure 26: TRACE model of LSTF

3-23
Figure 27: SNAP mask of TRACE model of LSTF

3-24
4 COMPARISON OF EXPERIMENTAL RESULTS

AND SIMULATIONS OF OECD/NEA ROSA

TEST 6.1

In a first step of the OECD/NEA ROSA test 6.1 simulation, the initial conditions were

achieved. The obtained values were in reasonable agreement with experimental values, as

shown in Table 6.

In a second step, several simulations with MODEL-1 (see page 3-10) were performed. Results

of these simulations agreed very well with experimental results, with deviation in some TH

parameters. Although phenomenology was well reproduced, a small delay in cladding and

core exit temperatures (Figures 29 and 30) was obtained.

Therefore, another model was developed including a nine U-tube heights in steam generators (MODEL-2, see page 3-10). Nevertheless, the simulations performed with MODEL-2

provided similar results than MODEL-1, see Figures 28 to 32 for more details.

In order to capture the delay in cladding and core exit temperatures new modifications were

tested (MODEL-3, see page 3-10). Firstly, several RFRIC factors were adjusted in the Utube and later the reflood model option was activated. These modifications diminished the

delay time between experimental and simulated data, as shown in Figure 33.

Finally, the test was simulated with the later and current TRACE model (MODEL-4, see

page 3-10), which gives better results, as shown in Figures 34 to 38. The primary and

secondary pressures matches fairly well with the experimental result. The evolution of the

core uncovery has the same behavior as in the test as well as the CET temperature. There

was only a small delay in primary pressure comparing to the test results. See Figures 34 to

38. As shown in Table 7, cronology of major events in the final TRACE simulation with

MODEL-4 matchs fairly well with experimental results.

Several snapshots of the Test 6.1 video obtained through SNAP with the LSTF mask and the

last TRACE simulation are shown in Figures 39 and 40, where void fraction is showed in bluewhite scale on plant components, green/red color indicates on/off status of pumps, valves

and main systems, and a red/blue scale shows the power of PZR heaters. Additionally,

primary/secondary pressure, pump rotational speed and cladding temperature profile are

depicted for each instant of the transient. The six pictures of those two snapshots correspond

to the main phases of the sequence:

1. Steady state,

2. Upper head break and depressurization in RCS with PZR empty and beginning of

4-1
voiding inside vessel,

3. Coastdown of reactor coolant pumps and emptying of SG U-tubes,

4. Beginning of core uncovery,

5. Maximum core uncovery and peak cladding temperature,

6. Accumulators discharge and core reflood (end of the simulation).

Additionally to the post-test simulations, a sensitivity analysis of RCP trip has been performed. In Test 6.1 the RCP trip occurs at the same time than the opening of break valve,

nevertherless the standard emergency operating procedures (EOP) of Westinghouse plants

(as the reference of LSTF), used in this kind of LOCA sequences (EOP E-1 and ES-1.2)

include two conditions in order to perform the RCP trip: loss of subcooling and availability

of HPSI. In the test there was not availability of HPSI and therefore the RCP would nor be

tripped in the case of EOP application. Next section present main conclusions and results

of the impact of RCP trip delay.

Items Specified Measured TRACE

(w/ (w/o) PZR) (w/ (w/o) PZR)

Pressure Vessel

Core power(MW) 10±0.07 10.12 10.00

Primary Loop

Hot Leg Fluid Temperature(K) 598.1±2.75 598.0/597.7 598.88/598.84

Cold Leg Fluid Temperature(K) 562.4±2.75 563.5/563.3 561.91/561.88

Mass Flow Rate (kg/s / loop) 24.3±1.25 24.9/24.88 23.91/23.77

Downcomer to Hot Leg Bypass (kg/s) 0.049±0.01 0.05/0.045 0.048/0.048

Pressurizer

Pressure(MPa) 15.5±0.108 15.51 15.5

Liquid Level(m) 7.2±0.25 7.18 7.30

Accumulator System

Pressure(MPa) 4.51±0.054 4.52/4.51 4.55/4.55

Temperature 320±2.3/2.4 321.6/321.9 320.0/320.0

Secondary loop

Secondary-Side Pressure (MPa) 7.3±0.054 7.33/7.33 7.09/7.08

Secondary-Side Liquid Level(m) 10.3±0.38 10.25/10.23 10.84/10.84

Steam Flow Rate (kg/s) 2.74±0.1 2.65/2.60 2.75/2.74

Main Feedwater Flow Rate (kg/s) 2.74±0.05 2.76/2.65 2.73/2.72

Main Feedwater Temperature(K) 495.2±2.63 495.9/495.1 495.2/495.2

Auxiliary Feedwater Temperature(K) 310±2.37 309.8 310/310

Table 6: Initial and boundary conditions for OECD/NEA ROSA test 6.1

4-2
EVENT Test 6.1 TIME (s) Simulation TIME (s)

Break valve open 2000 2000

SCRAM signal 2022 2019

SI signal 2028 2027

Primary coolant pumps stop 2276 2000

Primary pressure lower than SG ∼ 2800 ∼ 2600

secondary-side pressure

Core uncovery ∼ 2800 ∼ 2800

Initiation of SG secondary-side 3090 3190

depressurization

Core power decrease by LSTF ∼ 3200 ∼ 3400

core protection system

Initiation of accumulator system ∼ 3300 ∼ 3600

Initiation of LPSI system ∼ 4900 ∼ 5200

Table 7: Chronology of major events: comparison between OECD/NEA ROSA Test 6.1 and

TRACE simulation

2000 3000 4000 5000

Time (s)

0

5e+06

1e+07

1,5e+07

2e+07

pressure (Pa)

Trace 9 U-Tubes

Trace 1 U-Tube

Experimental

Figure 28: Primary pressure. Comparison between experimental and 1/9 U-tubes models

(MODEL-1 and MODEL-2)

4-3
2000 3000 4000 5000

Time (s)

400

600

800

1000

1200

max average rod temp (K)

Trace 1 U-Tube

Experimental

Trace 9 U-Tubes

Figure 29: Peak cladding temperature. Comparison between experimental and 1/9 U-tubes

models (MODEL-1 and MODEL-2)

2000 3000 4000 5000

Time (s)

400

500

600

700

800

gas temperature (K)

Trace 9 U-Tubes

Trace 1 U-Tube

Experimental

Figure 30: Core exit thermocouple temperature. Comparison between experimental and 1/9

U-tubes models (MODEL-1 and MODEL-2)

4-4
2000

Time (s)

-10

0

10

20

30

40

50

total mass-flow (kg/s)

Trace 9 U-tubes

Trace 1 U-Tube

Experimental

Figure 31: Loop B mass flow. Comparison between experimental and 1/9 U-tubes models

(MODEL-1 and MODEL-2)

2000 3000 4000 5000

Time (s)

0

1000

2000

3000

4000

5000

6000

7000

integ. mass flow (kg)

Trace 9 U-Tubes

Trace 1 U-Tube

Experimental

Figure 32: Break integrated mass flow. Comparison between experimental and 1/9 U-tubes

models (MODEL-1 and MODEL-2)

4-5
2000 2500 3000 3500 4000 4500

Time (s)

500

600

700

800

900

1000

max average rod temp (K)

JAEA

w/o rfric-reflood

w/ rfric-reflood

w/ rfric

Figure 33: Peak Cladding Temperature. Comparison between experimental data and models

with adjusted RFRIC and/or reflood option (MODEL-3)

Figure 34: PCT for test 6.1 (MODEL-4)

4-6
Figure 35: Primary and secondary pressures for test 6.1 (MODEL-4)

Figure 36: Core liquid level for test 6.1 (MODEL-4)

4-7
Figure 37: Break mass flow for test 6.1 (MODEL-4)

Figure 38: Temperature at core exit for test 6.1 (MODEL-4)

4-8
Figure 39: Void fraction in primary and secondary sides. Snapshots of SNAP video

simulation (1/2)

4-9
Figure 40: Void fraction in primary and secondary sides. Snapshots of SNAP video

simulation (2/2)

4-10
4.1 Sensitivity analysis to RCP trip delay

This section describes the main results and conclusins of a sensitivity analysis with repect

the instant of RCP trip, trying to identify the most penalizing conditions. Figures 42 to

47 include the results of simulations for different RCP trip times. In all cases, transient is

simulated with core protection system activated and deactivated. The sensitivity analisys

shows that the worst results are obtained for intermediate times of RCP trip, as shown in

Figure 47, where PCT is depicted as a function of RCP trip time (with and without core

protection system). These results prove that this task included in Westinghouse EOPs has a

great impact in the evolution of the transient. This issue is well known since TMI accident

due to several experimental tests and simulations that were performed for Westinghouse

reactors for SBLOCA sequences with and without availability of HPSI. The conclusion of

this sensitivity analysis is that if the objective of the experimental test is to analyze the

expected phenomenology in present Westinghouse nuclear power plants then it could be

interesting to perform a similar test without RCP trip. This kind of test could also be

interesting in order to analyze the behavior of the core exit thermocouple in sequences with

similar conditions to present Westinghouse EOPs.

So, in general, before conducting an experiment would be important to review the EOPs of

the different vendors that participate in the project in order to decide which management

actions are more adequate for implementing in the experiment. Particularly, the emergency

operating procedures (EOP) of Westinghouse used in LOCA sequences (EOP E-1 and ES-1.2,

see Figure 41) include two conditions in order to perform the RCP trip: loss of subcooling

and availability of HPSI.

Figure 41: Simplified scheme of standar Westinghouse EOPs in LOCA scenaries

4-11
0 1000 2000 3000 4000 5000

Time (s)

0

5e+06

1e+07

1,5e+07

2e+07

reactor power (W)

rpower-250-pump-stop-2000s-with-core-protection

rpower-250-pump-stop-2000s-without-core-protection

Figure 42: Core power with and without core protection system

2000 2500 3000 3500 4000 4500 5000

Time (s)

0

50

100

150

impeller rotational speed (rad/s)

pump-stop-3000s-without-core-protection

no-pump-stop-with-core-protection

pump-stop-2100s-without-core-protection

pump-stop-2000s-with-core-protection

pump-stop-2500s-without-core-protection

pump-stop-3250s-without-core-protection

pump-stop-2350s-without-core-protection

pump-stop-3500s-without-core-protection

pump-stop-2000s-without-core-protection

pump-stop-2750s-without-core-protection

pump-stop-2200s-without-core-protection

pump-stop-2060s-without-core-protection

Figure 43: Pump rotational speed for different RCP trip times

4-12
2000 2500 3000 3500 4000 4500

Time (s)

1000

2000

3000

4000

5000

6000

integ. mass flow (kg)

no-pump-stop-with-core-protection

pump-stop-2000s-with-core-protection

pump-stop-2000s-without-core-protection

no-pump-stop-without-core-protection

pump-stop-2200s-without-core-protection

pump-stop-2350s-without-core-protection

pump-stop-2500s-without-core-protection

pump-stop-2060s-without-core-protection

pump-stop-2750s-without-core-protection

pump-stop-2100s-without-core-protection

pump-stop-3000s-without-core-protection

pump-stop-3250s-without-core-protection

pump-stop-3500s-without-core-protection

Figure 44: Integrated break mass flow for different RCP trip times

2000 2500 3000 3500 4000 4500

Time (s)

1

2

3

sv 28: liq level (m)

no-pump-stop-with-core-protection

pump-stop-2000s-with-core-protection

pump-stop-2000s-without-core-protection

no-pump-stop-without-core-protection

pump-stop-2200s-without-core-protection

pump-stop-2350s-without-core-protection

pump-stop-2500s-without-core-protection

pump-stop-2060s-without-core-protection

pump-stop-2750s-without-core-protection

pump-stop-2100s-without-core-protection

pump-stop-3000s-without-core-protection

pump-stop-3250s-without-core-protection

pump-stop-3500s-without-core-protection

Figure 45: Core level for different RCP trip times

4-13
2800 3000 3200 3400 3600 3800 4000 4200

Time (s)

400

600

800

1000

1200

1400

Cladding temperature (K)

no-pump-stop-with-core-protection

pump-stop-2000s-with-core-protection

pump-stop-2000s-without-core-protection

no-pump-stop-without-core-protection

pump-stop-2200s-without-core-protection

pump-stop-2350s-without-core-protection

pump-stop-2500s-without-core-protection

pump-stop-2100s-without-core-protection

pump-stop-2060s-without-core-protection

pump-stop-2750s-without-core-protection

pump-stop-3000s-without-core-protection

pump-stop-3500s-without-core-protection

pump-stop-3250s-without-core-protection

Figure 46: PCT for different RCP trip times

Figure 47: PCT, with and without core protection system. Values at 5000 s represent no

RCP trip.

4-14
5 SIMULATION OF UPPER HEAD SBLOCA IN

ALMARAZ NPP

Almaraz NPP consists of two PWRs located in Caceres (Spain) and it is owned by a consortium of three spanish utilities: Iberdrola (53%), Endesa (36%) and Gas Natural Fenosa

(11%). Comertial operation started in April 1981 (Unit I) and September 1983 (Unit II).

Each unit has a PWR Westinghouse with three loops, as shown in Figure 48, and two turbines (high and low pressure turbines) in tandem compound. The nominal power is 2947

MWt for unit 1 and 2729 MWt for unit 2. It is equipped with three steam generators Siemens

KWU 61W/D3. Reactor coolant pumps are type single stage, centrifugal model W-11011-Al

(93-D) designed by Westinghouse. The AFWS consists of one turbine driven pump and two

motor driven pumps.

Relative positions and heights and diameters of RCS components are shown in Figures 48

and 49. In Table 8 can be read the main operating parameters for both units.

This section presents main simulation results of similar transients to test 6.1 with the TRACE

model of Almaraz NPP, taking into account different accident management actions and

conditions.

Figure 48: Scheme of Almaraz NPP (plan view)

5-1
Figure 49: Scheme of Almaraz NPP (elevation view)

5-2
Table 8: Main operating parameters of the Almaraz I NPP

Description Value

Thermal reactor power (Unit 1/Unit 2) 2947/2729 Mwt

Fuel UO2 + GdO2

Number of assemblies 157

Number of loops 3

Reactor operating pressure 155.017 bar

Coolant averaged temperature

Zero load 564.9 K

100 % 580.8 K

Steam generators Siemens KWU 61W/D3

Number of tubes (per SG) 5130

Total tube length (per SG) 108294.3 m

Tube inner diameter 17.96 mm

Tube material INCOLOY 800

Pumps type Centrifugal model W-11011-Al (93-D)

Pump discharge head 86.26 m

Design flow rate 6.27 m3/s

Pump speed 155.509 rad/s

Primary volumes

Vessel 100.81 m3

Hot leg (x3) 3.18 m3

Steam generator (x3) 32.28 m3

Cross leg (x3) 3.6 m3

Reactor coolant pump (x3) 4.02 m3

Cold leg (x3) 3.23 m3

Surge line 1.14 m3

Pressurizer 39.64 m3

Spray lines 0.45 m3

TOTAL 280.97 m3

Number of PZR relieve / safety valves 2/3

Number of PZR spray valves 2

Heaters capacity (proportional/backup) (377 kW / 1023 kW)

Maximum spray flow 0.022 m3/s·valve

Steam mass flow rate at 100%

SG1 489 kg/s

SG2 486 kg/s

SG3 500 kg/s

5-3
5.1 Description of Almaraz NPP model

Almaraz I NPP TRACE model has 255 thermal-hydraulic components (2 VESSEL, 73 PIPE,

43 TEE, 54 VALVE, 3 PUMP, 12 FILL, 33 BREAK, 32 HEAT STRUCTURE and 3 POWER

component), 740 SIGNAL VARIABLES, 1671 CONTROL BLOCKS and 58 TRIPS.

Figure 50 shows a schematic diagram of the TRACE model of Almaraz NPP. Regarding the

primary and secondary circuits, the following components have been modeled:

• Reactor vessel, modeled by a VESSEL component, Figure 51, which includes the core

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

via downcomer and via guide tubes.

• Nuclear core power is modeled with axial and radial cosine power shape distributions.

Core power is distributed into nine HEAT STRUCTURE components located each one

in one core sector.

• Primary circuit, including steam generators and pressurizer in loop 2 (containing

heaters, relief/safety valves and pressurizer spray system).

• Chemical and Volume Control System (CVCS).

• Emergency Core Cooling System (ECCS): safety injection system and accumulators

(ACCs).

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

valves, and the steam dump with the eight valves.

• FW and AFW systems. Feed water pumps coastdown and auxiliary mass flows are

included as boundary conditions.

5-4
Figure 50: TRACE model of Almaraz NPP (SNAP mask with mean components)

5-5
The control, protection and engineering safeguard systems and signals modeled are:

• Pressurizer level control:

– CVCS isolating discharge signal.

– CVCS charge flow and heaters.

• Pressurizer pressure control:

– Proportional and backup heaters.

– Spray lines.

– PORVs.

• Steam generators level control system.

• Steam dump control.

• Turbine control.

• Protection and engineering safeguard system-signals:

– Emergency shutdown system (SCRAM).

– Safety injection.

– Pressurizer safety valve logic.

– Auxiliary feedwater system activation and control

– Relief, safety and isolating valve logic of steam lines.

– Normal feedwater system isolation.

– Turbine trip.

– Pump trip.

This model has been validated with steady and transient conditions and verified with a large

set of transients. See References [1], [2], [3], [4], [5], [6] and [7].

5-6
5.2 Comparison between Almaraz NPP and LSTF

Several considerations had to be made in order to undertake the transposition to Almaraz

NPP of test 6.1 conditions, due to differences between the NPP and the facility. Tables 9

and 10 show main differences between LSTF, Tsuruga NPP and Almaraz NPP.

As can be seen in Figure 51, RPV differences between Almaraz NPP and LSTF are cause of

slightly differences in vessel models for both facilities. RPV models are quite similar, with

analogous modelling: both cores are nodalized with four radii and three sectors, although

Almaraz NPP core is nodalized with five more axial levels. The other components of the

vessel are modeled roughly equal in both cases: downcomer, lower and upper heads and

guide tubes show similar distribution and slight differences in nodalization.

LSTF TSU TSU/LSTF ALM ALM/LSTF

Primary Volume

(m3)

8.14 347 42.6 281 34.52

RPV total volume

(m3)

2.754 131.7 47.8 100.81 36.60

Upper head volume

(m3)

0.4963 24.6 49.6 11.81 23.80

Upper plenum volume (m3)0.4950 28.4 57.4 28 56.56

Core volume (m3) 0.4477 17.5 39.1 14.10 31.49

Lower plenum volume (m3)0.4644 29.62 63.8 20.20 43.47

Downcomer volume

(m3)

0.8504 31.58 37.1 20 23.52

Hot Leg area (m2) 0.03365 0.4261 12.66 0.42616393 12.66

Cold Leg area (m2) 0.03365 0.3831 11.39 0.38321039 11.39

Control rod drive

mechanism area (m2)

0.0006394 0.007280 11.38 0.003832 5.99

Instrumentation penetration area (m2)8.04E-6 5.07E-4 63 5.07E-4 63

Core Area (m2) 0.113 4.75 41.89 3.87 34.27

Downcomer Area

(m2)

0.086 3.38 39.39 2.53 29.41

Table 9: Comparison among LSTF, Almaraz NPP and Tsuruga NPP. Volumes and areas.

5-7
LSTF TSU TSU/LSTF ALM ALM/LSTF

Pressure (MPa) 16 16 1 16 1

Temperature (K) 598 598 1 598 1

Number of fuel rods 1008 50952 50.55 41448 41.12

Core height (m) 3.66 3.66 1 3.66 1

Power (MW) 10 3423 342 2686 268.6

Core inlet flow (m3/s) 0.0488 16.7 342 18.7 383

Number of loops 2 4 2 3 1.5

Number of U-tubes for SG 141 3382 24.0 5130 36.38

Mean U-tube length (m) 20.2 20.2 1 21.11 1.04

Hot Leg diameter (m) 0.207 0.737 3.56 0.73 3.55

Cold Leg diameter (m) 0.207 0.6985 3.374 0.698 3.372

Control rod drive mechanism.

Inner diameter (mm)

13.8 102 7.4 69.85 5.1

Instrumentation penetration.

Outer diameter (mm)

3.2 25.4 7.94 25.4 7.94

Hot Leg length (m) 3.69 6.99 1.89 7.25 1.96

Table 10: Comparison between LSTF, Almaraz NPP and Tsuruga NPP. TH parameters and

lengths

5-8
Figure 51: Comparison between Almaraz NPP and LSTF VESSEL components

5-9
5.3 Comparison between Almaraz NPP model and LSTF

experimental results

Several modifications in both sets of parameter models have attempted in order to fit results

obtained with Almaraz NPP model and LSTF. Section 5.4 will show further details of these

results.

In a first attempt, the upper head break area for LSTF was scaled-up for Almaraz model,

taking into account the volume scaling between vessel of LSTF and Tsuruga-2 NPP (1/48

volume ratio, see Table 10). Results show that the discharged coolant mass was greater than

the experimental one, as shown in Figure 52. Due to this difference the pressure decreases

quicker in Almaraz NPP model than in LSTF, see Figure 53 for more details.

In a second attempt, the break area was scaled in Almaraz NPP model taking into account

that the volume ratio between Almaraz NPP and LSTF is 1/34.52 instead of 1/48 (see Table

10). Transient and TH parameters result to be better adjusted than simulations with volume

factor 1/48 (scale factor between LSTF and Tsuruga-2), because scale factor of 1/34.52 is

more realistic for Almaraz NPP. Due to that, this volume factor was considered acceptable

for ongoing analisys. This volume factor was considered for subsequent analyses. Results

obtained for the integrated break flow is slightly lower for Almaraz NPP, as depicted in Figure

55. On the other hand, Figure 56 shows that the depressurization of the primary does not

occur at the same time, but with some delay, probably because less flow is discharged through

break. Peak cladding temperature rise in Almaraz NPP is lower and later than in LSTF (see

Figure 57).

Figure 52: Integrated break flow. Almaraz NPP vs 1/48 scaled LSTF

5-10
Figure 53: Primary and secondary pressure. Almaraz NPP vs 1/48 scaled LSTF

Figure 54: Core power comparison. Almaraz NPP vs 1/34.52 scaled LSTF

5-11
Figure 55: Integrated break flow. Almaraz NPP vs 1/34.52 scaled LSTF

Figure 56: Primary and secondary pressure. Almaraz NPP vs 1/34.52 scaled LSTF

5-12
Figure 57: Peak cladding temperature. Almaraz NPP vs 1/34.52 scaled LSTF

5-13
5.4 Sensitivity analysis in Almaraz NPP model

A broad sensitivity analysis has been also preformed in order to understand the differences

betrween LSTF and NPP transients. A broad spectrum of parametres have been explored

in order to find the most limiting cases.

1. Discharge coefficient (sensitivity examined after scaling LSTF model).

2. Break area.

3. Break localization in different radial sectors of the upper head model.

4. Friction factors in ACCs exit.

5. Initial accumulators pressure.

6. Accumulators flow (scaled from test ROSA 6.1).

7. Upper downcomer area.

8. Friction factors in guide tubes (FRIC and RFRIC).

Only results corresponding to relevant parameters are shown in the sequel.

As shown in Figure 55, flow discharged through break is slightly lower than in the case of

test 6.1. Discharge coefficients (chm12 and chm22 in TRACE input, for subcooled and

two-phase flow, respectively; ranging from 0.8 to 1.3) were adjusted to fit break discharge

flow; results are depicted in Figure 58. Discharge coefficients chm12 = 1.1 and chm22 =

1.2, allow to improve the starting time for rising of peak cladding temperature (see Figure

59), although are not able to capture experimental maximum value in PCT.

In a second stage, a sensitivity analisys of break area was performed, varying break area

between -10% and +75% of original break area1. Results obtained from simulations show

that there is a sligth dependence on break area of time when cladding temperature starts

to rise, but there is almost no dependence of maximum temperature reached by cladding

(see Figure 60, where integrated mass flow through break is depicted; and Figure 61, which

shows peak cladding temperature).

In a third stage, a modification in the model, varying break position in upper head was

carried out. UH break was varied from the center to the perimeter of the vessel, trying

1although several hypothetical break areas have been examined, it should be noted that the maximum

diameter of UH break in Almaraz NPP is only 6.99cm, according to size of control rod drive mechanism in

plant.

5-14
four different radial positions. Results of that sensitivity analisys are depicted in Figure 62,

where PCT is depicted for each radial position. Although there were not expected significant

differences between position, it can be noted that when simulation is carried out with UH

break near the perimeter of the vessel PCT maximum is much lesser than in other cases.

Figures 63 and 64 show that accumulators behaviour can explain this unexpected difference.

In all cases it can be noted very little dependence on position of integrated mass flow through

break.

In a fourth stage, different friction factors have been checked for accumulators outlet. As

shown in Figure 65, integrated discharged flow from accumulators is greatly modified through

variation of friction factor, being the model of accumulators outlet without friction the most

similar to experimental data of LSTF. The results show that peak cladding temperature has

very little dependence on accumulators outlet friction factor (see Figure 66).

Afterwards, the transient was simulated with two different downcomer areas: one with an

area 1/5 of original area and the other without upper downcomer area. Results of PCT are

depicted in Figure 67, where it can be noted that an area 1/5 times the original one has

very little differences in PCT. Only the case where upper downcomer is closed has significant

differences: in that case rise up time of cladding temperature is similar to experimental one,

as well as PCT maximum, which increases to almost 90% of PCT maximum in test 6.1, at the

same time. This modification causes another PCT local maximum, because of accumulators

discharge interruption in transient.

Figure 68 shows depressurization dependence on friction factors of vessel guide tubes. As

it can be noted, accumulators discharge is significantly affected by friction factors in guide

tubes, which causes meaningful differences in depressurization of RCS.

In summary, results of peak cladding temperature sensitivity cases show:

• Low sensitivity to break location, friction factors at ACCs exit and steady state

upper head mass flows.

• Medium sensitivity to discharge coefficients and upper downcomer area.

• High sensitivity with respect to break area size, RCP trip delay and number of ACCs

available.

For sensitivity analysis on other parameters, interested reader is referred to other works

of authors, e.g. Accident Management Actions in an Upper Head SBLOCA with HPSI Failed

[8], where a broad analysis of Upper Head SBLOCA with HPSI failed is performed taking

into account different accident management actions and conditions in order to check their

suitability.

5-15
Figure 58: Integrated break flow. Almaraz NPP vs 1/34.52 scaled LSTF with different

discharge coefficients

Figure 59: Peak Cladding Temperature. Almaraz NPP vs 1/34.52 scaled LSTF with different

discharge coefficients

5-16
Figure 60: Integrated break flow. Almaraz NPP vs 1/34.52 scaled LSTF with different break

areas

Figure 61: PCT sensitivity to break area

5-17
Figure 62: PCT sensitivity to break location in different radial sectors

Figure 63: Integrated discharged flow from accumulators. Almaraz NPP vs 1/34.52 scaled

LSTF. Cases with break in radial sectors 3 and 4

5-18
Figure 64: Primary and secondary pressures. Almaraz NPP vs 1/34.52 scaled LSTF with

upper break in radial sector number 4

Figure 65: Integrated discharged flow from accumulators. Almaraz NPP vs 1/34.52 scaled

LSTF. Cases with different friction factors at ACC outlet

5-19
Figure 66: PCT sensitivity to friction factors at ACC outlet

Figure 67: PCT sensitivity to friction Upper Downcomer Area

5-20
Figure 68: Primary and secondary pressures. Sensitivity to friction factors at guide tubes

5-21

6 EXECUTION STATISTICS

The simulations have been run in 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.

As shown in figures, CPU effort can be divided into two main stages: Figure 69 shows that

there is an increase in computing time at 2600 seconds of simulation, after coolant pumps are

tripped (2276 s) and before break flow becomes single-phase vapor (2700 s). On the other

hand, simulation with Almaraz model (Figure 70) shows an increase of computing time at

the same time the break occurs (4650 s). CPU time performance for both phases and in

both cases is summarized in Table 11.

Figure 69: CPU time performance for ROSA/LSTF test 6.1 simulation

Computing LSTF simulation Almaraz NPP simulation

Stage performance (%) performance (%)

Stage 45.4% 40.0%

1 (0-2600 s) (0-4650 s)

Stage 4.4% 4.0%

2 (2600-5000 s) (4650-80000 s)

Table 11: CPU performance for LSTF and Almaraz NPP simulations of test 6.1

6-1
0 1000 2000 3000 4000 5000 6000 7000

Time (s)

0

10000

20000

30000

40000

50000

60000

70000

80000

Total CPU time (s)

Figure 70: CPU time performance for Almaraz NPP model simulation of test 6.1

6-2
7 CONCLUSIONS

The Large Scale Test Facility replicates the primary and secondary coolant systems of a

Westinghouse 4-loop reactor of 3423 MWt, including ECCS and control systems for accident management actions. In OECD/NEA ROSA Test 6.1, phenomenology of upper head

SBLOCA with HPSI failed was analyzed, in order to obtain experimental thermal-hydraulic

data for the assessment of thermal-hydraulic computer codes like TRACE, and models for

plant integral analysis.

In this report, a post-test analisys of OECD/NEA ROSA Test 6.1 using TRACE 5.0 has

been presented, as well as the simulation of such a test in the TRACE model of a commercial

nuclear power plant (Almaraz NPP). Main findings were:

• Results of TRACE post-test for OECD/NEA ROSA Test 6.1 have been well reproduced

only through sucessive improvements and modifications in the LSTF model. Main

changes in order to obtain an adequate post-test simulation were RFRIC factors and

the activation of reflood model, as well as a renodalization of the vessel, cold and hot

legs and steam generators.

• Not all improvements added to the LSTF model have led to good results, contrary to

expectations. For example, the addition of 9 U-tubes in the SG of LSTF model yields

worse results than the simplest model, with only 1 U-tube.

• During OECD/NEA ROSA Test 6.1, one of the pre-established experimental conditions

was the coolant pump trip at the same time that break occurs. Sensitivity analisys

to RCP trip delay shows that maximum cladding temperature increases as RCP trip

delay increases, but only until approximately 1000 seconds after the break. Beyond

that point, PCT decreases slightly, remaining almost constant independently of RCP

trip delay.

• Simulation of OECD/NEA ROSA Test 6.1 in Almaraz NPP TRACE model requires

taking into account the scale factor between LSTF and Almaraz NPP. Considering

this scale factor, transient in both models seems very similar, with some deviations.

Depressurization is quite similar in both models until ACCs discharge, whose pressure

is higher in Almaraz NPP, and therefore discharge before. Cladding temperature begins

to rise up earlier in experimental test. Also, PCT is significantly lower in Almaraz NPP

simulations, due to ACCs discharge, which yields to core reflood.

• Simulation of Test 6.1 in Almaraz NPP model shows diverse grades of sensitivity to several parameter modifications. Break location, steady state upper head mass flow and

friction factors at ACCs exit lead to little change in results. Model exhibits medium

sensitivity to discharge coefficients and upper downcomer area, and shows high sensitivity to break area size, RCP trip delay and quantity of ACCs available.

7-1
As a general conclusion, it should be noted that TRACE code is adequate for simulating

upper head SBLOCA sequences.

7-2
8 REFERENCES

References

[1] Gonzalez,I.; Queral,C.; Exposito,A.; & Novales,M. TRACE Model of Almaraz Nuclear

Power Plant. In Nuclear Energy for New Europe, 2005.

[2] Lopez,A.; Queral,C. & Gallego,I. Conversion of Almaraz NPP model from RELAP5 into

TRAC-M. In ANS/ENS International Winter Meeting, 2003.

[3] Queral,C. et Al. Conversion of the thermal hydraulics components of Almaraz NPP

model from RELAP5 into TRAC-M. In Nuclear Energy for new Europe, 2002.

[4] Queral,C. et Al. Problems found in the conversion of Almaraz NPP model from RELAP5

into TRAC-M. Report send to nrc, DSE-UPM, Madrid, 2002.

[5] Queral,C. et Al. Conversion of the steam generator model of Almaraz NPP from RELAP5

into TRAC-M, TRAC-P and TRACE. In MC-05, Avignon, France, 2005.

[6] Queral,C. et Al. Verication and validation of Almaraz NPP TRACE model. In ICAPP

08, Anaheim, California, 2008.

[7] Queral,C. et Al. NUREG/IA-0233. Assessment of TRACE 4.160 and 5.0 against RCP

Trip Transient in Almaraz I Nuclear Power Plant. DSE - UPM, July 2010.

[8] Queral,C. et Al. Accident Management Actions in an Upper Head SBLOCA with HPSI

Failed. Nuclear Technology, September 2011.

8-1




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NUCLEAR REGULATORY COMMISSION

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NUREG/IA-0426 Simulation of LSTF Upper Head Break (OECD/NEA ROSA Test 6.1) with 

TRACE Code. Application to a PWR NPP Model