ISBN 92-64-01064-5
NEA/NSC/DOC(2004)21

NEA NUCLEAR SCIENCE COMMITTEENEA COMMITTEE ON SAFETY OF NUCLEAR INSTALLATIONSBOILING WATER REACTOR
TURBINE TRIP (TT) BENCHMARK
Volume II: Summary Results of Exercise 1

June 2005

byBedirhan Akdeniz and Kostadin N. Ivanov
Nuclear Engineering Program
The Pennsylvania State University
University Park, PA, 16802 USA
Andy M. Olson
Exelon Nuclear
200 Exelon Way, KSA2-N
Kennett Square, PA, 19348 USA© OECD 2005
NEA No. 4448

NUCLEAR ENERGY AGENCY
ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT
AND US NUCLEAR REGULATORY COMMISSIONORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 30 democracies work together to address the economic, socialand environmental challenges of globalisation. The OECD is also at the forefront of efforts to understand and to help
governments respond to new developments and concerns, such as corporate governance, the information economy and thechallenges of an ageing population. The Organisation provides a setting where governments can compare policy experiences, seek
answers to common problems, identify good practice and work to co-ordinate domestic and international policies.
The OECD member countries are: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France,
Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Korea, Luxembourg, Mexico, the Netherlands, New Zealand, Norway,
Poland, Portugal, the Slovak Republic, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The
Commission of the European Communities takes part in the work of the OECD.
OECD Publishing disseminates widely the results of the Organisation’s statistics gathering and research on economic,
social and environmental issues, as well as the conventions, guidelines and standards agreed by its members.
***This work is published on the responsibility of the Secretary-General of the OECD. The opinions expressed andarguments employed herein do not necessarily reflect the official views of the Organisation or of the governments of its member
countries.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC European
Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its first non-European full
member. NEA membership today consists of 28 OECD member countries: Australia, Austria, Belgium, Canada, the Czech
Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, theNetherlands, Norway, Portugal, Republic of Korea, the Slovak Republic, Spain, Sweden, Switzerland, Turkey, the United
Kingdom and the United States. The Commission of the European Communities also takes part in the work of the Agency.
The mission of the NEA is:

to assist its member countries in maintaining and further developing, through international co-operation, thescientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclear
energy for peaceful purposes, as well as
to provide authoritative assessments and to forge common understandings on key issues, as input to government
decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainable
development.

Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive waste
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law andliability, and public information. The NEA Data Bank provides nuclear data and computer program services for participating
countries.
In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency in Vienna,
with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.
© OECD 2005
No reproduction, copy, transmission or translation of this publication may be made without written permission. Applications
should be sent to OECD Publishing: rights@oecd.org or by fax (+33-1) 45 24 13 91. Permission to photocopy a portion of thiswork should be addressed to the Centre Français d’exploitation du droit de Copie, 20 rue des Grands Augustins, 75006 Paris,
France (contact@cfcopies.com).

FOREWORDThe OECD Nuclear Energy Agency (NEA) completed under US Nuclear Regulatory Commission(NRC) sponsorship a PWR main steam line break (MSLB) benchmark against coupled systemthree-dimensional (3-D) neutron kinetics and thermal-hydraulic codes. Another OECD/NRC
coupled-code benchmark was recently completed for a BWR turbine trip (TT) transient and is theobject of the present report.
Turbine trip transients in a BWR are pressurisation events in which the coupling between core
space-dependent neutronic phenomena and system dynamics plays an important role. The data madeavailable from actual experiments carried out at the Peach Bottom 2 plant make the present benchmark
particularly valuable. While defining and co-ordinating the BWR TT benchmark, a systematic
approach and level methodology not only allowed for a consistent and comprehensive validation
process, but also contributed to the study of key parameters of pressurisation transients. The benchmark
consists of three separate exercises, two initial states and five transient scenarios.
The BWR TT Benchmark will be published in four volumes as NEA reports. CD-ROMs will also
be prepared and will include the four reports and the transient boundary conditions, decay heat values asa function of time, cross-section libraries and supplementary tables and graphs not published in thepaper version. BWR TT Benchmark – Volume I: Final Specifications was issued in 2001
[NEA/NSC/DOC(2001)1]. The benchmark team [Pennsylvania State University (PSU) in co-operation
with Exelon Nuclear and the NEA] has been responsible for co-ordinating benchmark activities,
answering participant questions and assisting them in developing their models, as well as analysing
submitted solutions and providing reports summarising the results for each phase. The benchmark
team has also been involved in the technical aspects of the benchmark, including sensitivity studies forthe different exercises. In performing these tasks, the PSU team has been collaborating with Andy M.
Olson and Kenneth W. Hunt of Exelon Nuclear. Lance J. Agee, of the Electric Power ResearchInstitute (EPRI), has also provided technical assistance for this international benchmark project.
Volume II summarises the results for Exercise 1 of the benchmark and identifies the key parameters
and important issues concerning the thermal-hydraulic system modelling of the TT transient withspecified core average axial power distribution and fission power (or reactivity) time transient history.
Exercise 1 helped the participants initialise and test their system code models for further use in
Exercise 3 on coupled 3-D kinetics/system thermal-hydraulics simulations.
Readers are invited to note that many of the original graphics in the report are in colour; colour
versions are available online at the NEA website (www.nea.fr).

3

Acknowledgements

The authors would like to thank Professor J. Aragonés of UPM, Dr. T. Lefvert and
Dr. S. Langenbuch of GRS, and Dr. F. Eltawila of NRC, whose support and encouragement inestablishing this benchmark were invaluable.
This report is the sum of many efforts – the participants and the funding agencies and their staff,
including the US Nuclear Regulatory Commission and the Organisation of Economic Co-operation
and Development. Special appreciation is due to: L. Agee of EPRI, Professor T. Downar of Purdue
University, B. Aktas of ISL Inc., Dr. G. Gose and Dr. C. Peterson from CSA, Dr. A. Hotta of TSI,
Dr. P. Coddington of PSI and Dr. U. Grundmann of FZR. Their technical assistance, comments andsuggestions were very valuable. We would like to thank them for the effort and time involved.
Of particular note are the labours of Dr. F. Eltawila, assisted by Dr. J. Han and Dr. J. Uhle of the
US Nuclear Regulatory Commission. Through their endeavours, funding was secured for this project.
We also thank them for their invaluable technical advice and assistance.
The authors wish to express their sincere appreciation for the outstanding support offered by
Dr. E. Sartori, who provides efficient administration, organisation and valuable technical advice.
The authors would also like to particularly thank all of the OECD/NEA BWR Turbine Trip
Benchmark participants for their valuable support, comment and feedback during this study.
Special thanks are due to the authors’ family members for their encouragement, support andpatience throughout the course of this work.
Finally, the authors are grateful to A. Griffin-Chahid for having devoted her competence andskills to the editing of this report.

4

TABLE OF CONTENTS

Foreword ........................................................................................................................................

3

List of Figures ................................................................................................................................

7

List of Tables..................................................................................................................................

9

List of Abbreviations.....................................................................................................................

11

Chapter 1 INTRODUCTION.....................................................................................................

13

Chapter 2 DESCRIPTION OF EXERCISE 1..........................................................................

15

2.1 General ................................................................................................................

15

2.2 Description of turbine trip scenario.....................................................................

15

2.3 Initial steady state conditions ..............................................................................

16

2.4 Transient calculations..........................................................................................

18

Chapter 3 METHODOLOGIES FOR COMPARATIVE ANALYSIS..................................

19

3.1 Standard techniques for the comparison of results..............................................

19

3.1.1 Integral parameter values..........................................................................

19

3.1.2 One-dimensional (1-D) axial distributions ...............................................

20

3.1.3 Time histories............................................................................................

20

3.2 ACAP analysis ....................................................................................................

21

Chapter 4 RESULTS AND DISCUSSION OF EXERCISE 1 ................................................

27

Chapter 5 CONCLUSIONS .......................................................................................................

65

References ......................................................................................................................................

67

Appendix A Description of computer codes used for analysis of Exercise 1of the NEA-NRC BWR TT Benchmark...................................................................

69

CATHARE (CEA, France).......................................................................................

71

S-RELAP5 (FANP, Germany) .................................................................................

73

ATHLET (GRS, Germany) ......................................................................................

73

DNB/3D (NETCORP, USA)....................................................................................

73

TRAC-BF1 (NFI, Japan)..........................................................................................

74

5

TRAC-BF1 (NUPEC, Japan) ...................................................................................

74

RETRAN-3D (PSI, Switzerland) .............................................................................

74

TRAC-BF1/MOD1 (PSU, USA)..............................................................................

75

TRAC-M (PSU/NRC, USA) ....................................................................................

77

TRAC-BF1-ENTRÉE (TEPSYS, Japan) .................................................................

78

RELAP5 (UPISA, Italy)...........................................................................................

80

TRAC-BF1-VALKIN (UPV, Spain)........................................................................

80

POLCA-T (Westinghouse, Sweden) ........................................................................

81

Appendix B Questionnaire for Exercise 1 of the NEA-NRC BWR TT Benchmark....................

85

CATHARE2 v1.5 (CEA, France).............................................................................

88

S-RELAP5 (FANP, Germany) .................................................................................

91

ATHLET (GRS, Germany) ......................................................................................

94

DNB/3D (NETCORP, USA)....................................................................................

97

TRAC-BF1 (NFI, Japan)..........................................................................................

103

TRAC-BF1 (NUPEC, Japan) ...................................................................................

106

RETRAN-3D (PSI, Switzerland) .............................................................................

109

TRAC-BF1 (PSU, USA) ..........................................................................................

116

TRAC-M (PSU/NRC, USA) ....................................................................................

118

TRAC-BF1 (TEPSYS, Japan)..................................................................................

121

RELAP5 (UPISA, Italy)...........................................................................................

123

TRAC-BF1-VALKIN (UPV, Spain)........................................................................

127

POLCA-T (Westinghouse, Sweden) ........................................................................

130

6

List of Figures
Figure 2.3-1.

PB2 TT2 initial core axial average power distribution from P1 edit .....................

17

Figure 3.2-1.

FOM configuration in ACAP ................................................................................

22

Figure 4-1.

Core inlet enthalpy.................................................................................................

30

Figure 4-2.

Core pressure drop.................................................................................................

31

Figure 4-3.

Core average void fraction distribution .................................................................

32

Figure 4-4.

Core average void fraction distribution .................................................................

32

Figure 4-5.

Void fraction distribution – averaged vs. Exelon data...........................................

33

Figure 4-6.

Void fraction distribution – Exelon and deviation.................................................

33

Figure 4-7.

Vessel dome pressure (5 seconds) .........................................................................

36

Figure 4-8.

Vessel dome pressure (5 seconds) .........................................................................

36

Figure 4-9.

Vessel dome pressure – measured vs. averaged data (5 seconds) .........................

37

Figure 4-10.

Vessel dome pressure (1.5 seconds) ......................................................................

38

Figure 4-11.

Vessel dome pressure (1.5 seconds) ......................................................................

38

Figure 4-12.

Vessel dome pressure – measured vs. averaged data (1.5 seconds) ......................

39

Figure 4-13.

Core exit pressure (5 seconds) ...............................................................................

40

Figure 4-14.

Core exit pressure (5 seconds) ...............................................................................

40

Figure 4-15.

Core exit pressure – Exelon vs. averaged data (5 seconds) ...................................

41

Figure 4-16.

Core exit pressure (1.5 seconds) ............................................................................

42

Figure 4-17.

Core exit pressure (1.5 seconds) ............................................................................

42

Figure 4-18.

Core exit pressure – Exelon vs. averaged data (1.5 seconds) ................................

43

Figure 4-19.

Jet pump flow rate (5 seconds) ..............................................................................

44

Figure 4-20.

Jet pump flow rate (5 seconds) ..............................................................................

44

Figure 4-21.

Jet pump flow rate – Exelon vs. averaged data (5 seconds) ..................................

45

Figure 4-22.

Jet pump flow rate (1.5 seconds) ...........................................................................

46

Figure 4-23.

Jet pump flow rate (1.5 seconds) ...........................................................................

46

7

Figure 4-24.

Jet pump flow rate – Exelon vs. averaged data (5 seconds) ..................................

47

Figure 4-25.

Core in-channel flow rate (5 seconds) ...................................................................

48

Figure 4-26.

Core in-channel flow rate (5 seconds) ...................................................................

48

Figure 4-27.

Core in-channel flow rate – Exelon vs. averaged data (5 seconds) .......................

49

Figure 4-28.

Core in-channel flow rate (1.5 seconds) ................................................................

50

Figure 4-29.

Core in-channel flow rate (1.5 seconds) ................................................................

50

Figure 4-30.

Core in-channel flow rate – Exelon vs. averaged data (1.5 seconds) ....................

51

Figure 4-31.

Steam bypass flow rate (5 seconds.)......................................................................

52

Figure 4-32.

Steam bypass flow rate (5 seconds).......................................................................

52

Figure 4-33.

Steam bypass flow rate – Exelon vs. averaged data (5 seconds) ...........................

53

Figure 4-34.

Steam bypass flow rate (1.5 seconds)....................................................................

54

Figure 4-35.

Steam bypass flow rate (1.5 seconds)....................................................................

54

Figure 4-36.

Steam bypass flow rate – Exelon vs. averaged data (1.5 seconds) ........................

55

Figure 4-37.

Turbine inlet pressure (5 seconds) .........................................................................

56

Figure 4-38.

Turbine inlet pressure (5 seconds) .........................................................................

56

Figure 4-39.

Turbine inlet pressure (1.5 seconds) ......................................................................

58

Figure 4-40.

Turbine inlet pressure (1.5 seconds) ......................................................................

58

Figure 4-41.

RETRAN PB2 thermal-hydraulic model...............................................................

61

Figure A-1.

TEPSYS modifications in control block INTM.....................................................

79

Figure B-1.

CEA nodalisation diagram.....................................................................................

88

Figure B-2.

FANP nodalisation diagram...................................................................................

91

Figure B-3.

GRS reactor pressure vessel nodalisation schema .................................................

94

Figure B-4.

GRS steam line nodalisation schema.....................................................................

95

Figure B-5.

NETCORP reactor vessel nodalisation..................................................................

99

Figure B-6.

NETCORP schematic of BWR NSSS representation............................................

99

Figure B-7.

NETCORP’s method of characteristics of main system........................................

1008

Figure B-8.

NETCORP schematic of reactor vessel geometry for jet pumps...........................

101

Figure B-9.

NETCORP flow diagram within reactor vessel for jet pumps ..............................

102

Figure B-10.

NETCORP DNB/3D equations .............................................................................

102

Figure B-11.

NFI TRAC-BF1 nodalisation diagram ..................................................................

103

Figure B-12.

NFI TRAC-BF1 steam line nodalisation ...............................................................

104

Figure B-13.

NUPEC Peach Bottom 2 plant nodalisation ..........................................................

106

Figure B-14.

PSI nodalisation of the reactor vessel ....................................................................

110

Figure B-15.

PSI nodalisation of the channels............................................................................

111

Figure B-16.

PSI nodalisation of the steam line..........................................................................

112

Figure B-17.

Nodalisation of the steam separator region............................................................

114

Figure B-18.

PSU TRAC-BF1 nodalisation................................................................................

116

Figure B-19.

PSU/NRC TRAC-M nodalisation..........................................................................

119

Figure B-20.

PSU/NRC TRAC-M jet pump nodalisation...........................................................

120

Figure B-21.

PSU/NRC TRAC-M separator nodalisation ..........................................................

120

Figure B-22.

TEPSYS TRAC-BF1 plant model .........................................................................

121

Figure B-23.

UPISA RELAP5 model .........................................................................................

124

Figure B-24.

UPISA steam line simplified schema ....................................................................

124

Figure B-25.

UPV thermal-hydraulic nodalisation .....................................................................

127

Figure B-26.

UPV thermal-hydraulic channel map ....................................................................

128

Figure B-27.

UPV steam line nodalisation..................................................................................

128

Table 1-1.

List of participants in Exercise 1 of the BWR TT benchmark...............................

14

Table 2.3-1.

PB2 TT2 initial conditions from process computer P1 edit...................................

16

Table 2.3-2.

PB2 TT2 initial core average axial power distribution from P1 edit .....................

17

Table 2.4-1.

PB TT2 event timing (time in milliseconds)..........................................................

18

Table 4-1.

Participant models used in Exercise 1 of the BWR TT benchmark.......................

29

List of Tables

9Table 4-2.

Core inlet enthalpy – deviation (ei) and FOM .......................................................

30

Table 4-3.

Core pressure drop – deviation (ei) and FOM........................................................

31

Table 4-4.

Core average void fraction – Exelon and deviation...............................................

34

Table 4-5.

Core average void fraction FOM ...........................................................................

35

Table 4-6.

Vessel dome pressure FOM – participant vs. measured data (5 seconds) .............

37

Table 4-7.

Vessel dome pressure FOM – participant vs. measured data (1.5 seconds) ..........

39

Table 4-8.

Core exit pressure FOM – participant vs. Exelon data (5 seconds).......................

41

Table 4-9.

Core exit pressure FOM – participant vs. Exelon data (1.5 seconds)....................

43

Table 4-10.

Jet pump flow rate FOM – participant vs. Exelon data (5 seconds) ......................

45

Table 4-11.

Jet pump flow rate FOM – participant vs. Exelon data (1.5 seconds) ...................

47

Table 4-12.

Core in-channel flow rate FOM – participant vs. Exelon data (5 seconds) ...........

49

Table 4-13.

Core in-channel flow rate FOM – participant vs. Exelon data (1.5 seconds) ........

51

Table 4-14.

Steam bypass flow rate FOM – participant vs. Exelon data (5 seconds)...............

53

Table 4-15.

Steam bypass flow rate FOM – participant vs. Exelon data (1.5 seconds)............

55

Table 4-16.

Turbine inlet pressure FOM – participant vs. averaged data (5 seconds)..............

57

Table 4-17.

Turbine inlet pressure FOM – participant vs. averaged data (1.5 seconds)...........

57

Table 4-18.

Sequence of events ................................................................................................

59

Table 4-19.

Time of the maximum peak ...................................................................................

59

Table 4-20.

Maximum peak value (delta changes) ...................................................................

60

Table A-1.

TEPSYS reference numerical options in ENTRÉE and valve parameters ............

79

Table B-1.

UPISA steam line and bypass system geometrical data ........................................

123

Table B-2.

UPISA decay constant and fractions of delayed neutrons .....................................

125

10

List of Abbreviations

1-D
2-D
3-D
ACAP
APRM
AVZ
BC
BE
BOC
BP
BPV
BVBO
BVFO
BWR
CA
CEA
CEPIR
DPIR
EOC
EOT
EPRI
Exelon
FA
FANP
FFT
FOM
FZR
GE
GRS
HFP
HZP
LPRM
LWR
ME
MSIV
MSLB
MVCEP
NEA

One-Dimensional
Two-Dimensional
Three-Dimensional
Automated Code Assessment Program
Average Power Range Monitor
Above Vessel ZeroBoundary Conditions
British Energy
Beginning of Cycle
Burnable Poison
Bypass Valve
Bypass Valve Begins Opening
Bypass Valve Full Open
Boiling Water Reactor
Control AssemblyCommissariat à l’Énergie AtomiqueCore Exit Pressure Initial Response
Dome Pressure Initial Response
End of Cycle
End of TransientElectric Power Research InstituteExelon Nuclear
Fuel Assembly
FRAMATOME ANP
Fast Fourier Transform
Figure of Merit
Forschungs Zentrum Rossendorf
General Electric
Gesellschaft für Anlagen- und Reaktorsicherheit mbH
Hot Full PowerHot Zero Power
Local Power Range Monitor
Light Water Reactor
Mean Error
Main Steam Isolation Valve
Main Steam Line Break
Maximum Value Percent Initial Value of Core Exit Pressure
Nuclear Energy Agency
11

NEM
NETCORP
NFI
NP
NPP
NRC
NRS
NSC
NSSS
NUPEC
OECD
PB
PB2
PECo
PSI
PSU
PWR
SRV
TEPSYS
T-H
TSV
TSVC
TT
TT2
UPISA
UPV
VBA
WES

Nodal Expansion Method
Nuclear Engineering Technology Corporation
Nuclear Fuel Industries Ltd.
Normalised Power
Nuclear Power Plant
US Nuclear Regulatory Commission
Nuclear Reactor Systems
Nuclear Science Committee
Nuclear Steam Supply System
Nuclear Power Engineering Corporation
Organisation for Economic Co-operation and Development
Peach Bottom Atomic Power Station
Peach Bottom Atomic Power Station Unit 2
Philadelphia Electric Company
Paul Scherrer InstitutePennsylvania State University
Pressurised Water Reactor
Safety Relief ValveTEPCO Systems Corporation
Thermal-Hydraulic
Turbine Stop Valve
Turbine Stop Valve Closing
Turbine Trip
Turbine Trip Test 2
University of Pisa
Universidad Politecnica de Valencia
Visual Basic for Applications
Westinghouse Atom AB

12Chapter 1
INTRODUCTION

Prediction of a nuclear power plant’s behaviour under both normal and abnormal conditions has
important ramifications for safety and economic operations. Such prediction is only possible using
highly sophisticated computer codes given the complexity of nuclear power plants. Incorporation offull three-dimensional (3-D) models of the reactor core into system transient codes allows for a“best-estimate” calculation of interactions between core behaviour and plant dynamics. Recent progressin computer technology has made the development of such coupled code systems (thermal-hydraulicand neutron kinetics) feasible. Considerable effort has been made by various countries and organisationsin this direction. In order to verify the capability of the coupled codes for the analysis of complex
transients with coupled core-plant interactions and to fully test thermal-hydraulic coupling, appropriate
light water reactor (LWR) transient benchmarks need to be developed on a higher best-estimate level.
Previous sets of transient benchmark problems separately addressed system transients (designed
mainly for thermal-hydraulic system codes with point kinetics models) and core transients (designed
for thermal-hydraulic core boundary condition models coupled with a 3-D neutron kinetics model).
The Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development(OECD) has recently completed, under the sponsorship of the US Nuclear Regulatory Commission(NRC), a PWR Main Steam Line Break (MSLB) benchmark [1] against coupled thermal-hydraulicand neutron kinetics codes.
A benchmark team from The Pennsylvania State University (PSU) was responsible for developing
the benchmark specifications, assisting the participants and co-ordinating the benchmark activities.
The benchmark was well-received by the international community. The participants of the PWRMSLB benchmark felt that there should be a similar benchmark against the codes for a BWR planttransient. A turbine trip (TT) transient in a BWR is a pressurisation event in which the coupling
between core phenomena and system dynamics plays an important role. In addition, the available realplant experimental data [3-5] makes such a benchmark problem quite valuable. The NEA, OECD and
US NRC have approved the BWR TT benchmark for the purpose of validating advanced-systembest-estimate analysis codes.
As a result, a benchmark project has been established in order to test the coupled system(thermal-hydraulic/neutron kinetics) codes by modelling a PB2 (a GE-designed BWR/4) turbine triptransient with a sudden closure of the turbine stop valve. Three turbine trip transients at different
power levels were performed at the Peach Bottom (PB2) BWR/4 nuclear power plant (NPP) prior to
shutdown for refuelling at the end of Cycle 2 in April 1977. The second test was selected for thebenchmark to investigate the effect of the pressurisation transient (following the sudden closure of theturbine stop valve) on the neutron flux in the reactor core. In a best-estimate manner the test conditions
approached the design basis conditions as much as possible. The actual data were collected, including
a compilation of reactor design and operating data for Cycles 1 and 2 of PB, as well as the planttransient experimental data. The transient was selected for this benchmark because it is a dynamicallycomplex event for which neutron kinetics in the core were coupled with thermal-hydraulics in thereactor primary system.

13

The purposes of this benchmark are met through the application of three exercises, which aredescribed in Volume 1 of the OECD/NRC BWR TT Benchmark – Final Specifications [2]. The purposeof the first exercise is to test the thermal-hydraulic system response and to initialise the participants’system models. Core power response is provided to reproduce the actual test results, using either
power or reactivity versus time data. The second exercise has two steady state conditions – hot zero
power (HZP) conditions and initial conditions of TT2. The purpose of Exercise 2 is to provide a clean
initialisation of the coupled core models since the core thermal-hydraulic boundary conditions areprovided. The last exercise, Exercise 3, is the best estimate for coupled 3-D core/thermal-hydraulicsystem modelling. This exercise combines elements of the first two exercises of this benchmark andis an analysis of the transient in its entirety. Exercise 3 also includes extreme scenarios, which provide
the opportunity to better test the code coupling and feedback modelling.
The following chapters are based on the comparative analysis of the submitted results for
Exercise 1. In total, fourteen results were submitted by participants representing fourteen organisations
from eight countries. A list of participants, who submitted results to the PSU benchmark team for the
first exercise, along with the codes used to perform the analysis, is located in Table 1-1. A moredetailed description of each code is presented in Appendix A and the modelling assumptions made byeach participant are given in Appendix B.
Chapter 2 contains a description of Exercise 1, including the initial conditions. Chapter 3
discusses the utilised comparative statistical methodology for integral parameters, one-dimensional(1-D) values and time histories. Chapter 4 provides a comparative analysis of the final results for thisfirst exercise. Finally, Chapter 5 provides a brief summary of the conclusions drawn from the analysis
of Exercise 1.
Table 1-1. List of participants in Exercise 1 of the BWR TT benchmark
Participant
number
1
2
3
4
5
6
7
8
9
10
11
12
13
14

Company name
CEA
ExelonFANPGRSNETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

Code

Country

CATHARE V15A/Mod2.1
RETRANS-RELAP5 V3.1.1
ATHLET
DNB-3D
TRAC-BF1
TRAC-BF1
RETRAN-3D
TRAC-BF1
TRAC-M
TRAC-BF1
RELAP5/Mod3.3
TRAC-BF1
POLCA-T

FranceUSA
Germany
Germany
USA
Japan
Japan
Switzerland
USA
USA
Japan
Italy
Spain
Sweden14

Chapter 2
DESCRIPTION OF EXERCISE 1

2.1 General
The first exercise consists of performing a system thermal-hydraulic calculation (no neutron
kinetics model involved; core power or reactivity as a function of time is provided as an input) for the
PB2 TT transient. The purpose of the first exercise is to test the thermal-hydraulic system response andto initialise the participants’ system models. This approach is used to develop/refine key sub-modelsof steam lines, steam bypass systems, jet pumps, steam separators and the upper downcomer region
where a distinct steam-water interface exists.
2.2 Description of turbine trip scenario
The Peach Bottom Unit 2 Turbine Trip Test 2 starts with a sudden closure of the turbine stop
valve (TSV) followed by the opening of the turbine bypass valve. From a fluid-flow phenomena pointof view, pressure and flow waves play an important role during the early phase of the transient. This isbecause rapid valve actions cause sonic waves, as well as secondary waves, which are generated in thepressure vessel. The pressure oscillation generated in the main steam piping propagates with relatively
little attenuation into the reactor core. The induced core pressure oscillation results in changes to thecore void distribution and fluid flow. The magnitude of the neutron flux transient in the BWR core isaffected by the initial rate of pressure rise (caused by pressure oscillation) and has spatial variation.
The simulation of the power response to the pressure pulse and subsequent void collapse requires a
3-D core modelling supplemented by a 1-D simulation of the remainder of the reactor coolant system.
A TT transient in a BWR-type reactor is considered one of the most complex events to be analysedbecause it involves the reactor core, the high pressure coolant boundary, associated valves and piping
in highly complex interactions with rapidly changing variables.
As mentioned earlier, the transient begins with a sudden TSV closure that initiates a pressure
wave in the main steam system, which is quickly transmitted to the reactor pressure vessel. While theTSVs are closing, the bypass system valves are designed to open, which allows for steam release and,thus, pressure relief. Safety relief valves (SRVs) begin to open at pre-established set points, providing
additional pressure relief. The pressure wave requires a detailed nodalisation modelling of the steam
system and its associated valves to assure that timing effects and pressure wave magnitude can be
accurately determined. This assures the availability of a pressure history on each valve, allowing
adequate modelling of steam flow through the valves.

15

2.3 Initial steady state conditions
The reference design for the BWR is derived from real reactor, plant and operation data for the
PB2 NPP and it is based on the information provided in the EPRI reports [3-6] as well as some
additional sources such as the PECo Energy Topical report [7].
Peach Bottom Atomic Power Station Unit 2 is a GE-designed BWR/4 with a rated thermal powerof 3 293 MW, a rated core flow of 12 915 kg/s (102.5 × 106 lb/hr), a rated steam flow of 1 685 kg/s
(13.37 × 106 lb/hr) and a turbine inlet pressure of 6.65 MPa (965 psia). The nuclear steam supply
system (NSSS) has turbine-driven feed pumps and a two-loop M-G driven recirculation system,
feeding a total of 20 jet pumps. In total, there are four steam lines and each has a flow-limiting nozzle,
main steam isolation valves (MSIVs), safety relief valves and a turbine stop valve. The steam bypass
system consists of nine bypass valves (BPVs) mounted on a common header, which is connected toeach of the four steam lines.
There are 764 fuel bundles with an active fuel length of 365.76 cm (12 ft) in the core region. The
fuel bundles consist of 576 original 7 × 7 fuel bundles with P/D (pitch/outer diameter) equal to1.87452 cm/1.43002 cm (0.738 in/0.563 in) and 188 partially reloaded 8 × 8 fuel bundles with P/Dequal to 1.62560 cm/1.25222 cm (0.640 in/0.493 in). Additionally, the core region includes 185 control
rods (CRs). For the reactor protection system (RPS), control systems for reactor pressure, recirculation
flow, feedwater flow and reactor water level are commonly used in reactors of this design.
Turbine trip test 2 (TT2) was initiated from steady state conditions after obtaining P1 edits fromthe process computer for nuclear and thermal-hydraulic conditions of the core. PB2 was chosen for theturbine trip tests because it is a large BWR/4 with relatively small turbine bypass capacity. At the
beginning of the test, the initial thermal power level was 61.6% of nominal power (2 030 MW); core
flow was 80.9% of nominal flow rate (10 445 kg/s, 82.9 × 106 lb/hr); and average power range monitor(APRM) scram set point was 95% of nominal power. (PBTT benchmark participants used the control
rod insertion time of 0.75 seconds [s] in all exercises.) For the TT2 test, the dynamic measurements
were taken with a high-speed digital data acquisition system capable of sampling over 150 signals
every 6 milliseconds (ms) and the core power distribution measurements were taken from the plant’s
local in-core flux detectors. Special fast-response pressure and differential pressure transducers were
installed in parallel with the existing plant instruments in the nuclear steam supply system. Table 2.3-1
provides the reactor initial conditions for performing steady state calculations.
Table 2.3-1. PB2 TT2 initial conditions from process computer P1 edit
Core thermal power (MWt)
Initial power level (% of rated)
Gross power output (MWe)Feedwater flow (kg/s)
Reactor pressure (Pa)
Total core flow (kg/s)
Core inlet sub-cooling (J/kg)
Feedwater temperature (K)Core pressure drop (Pa)Jet pump driving flow (kg/s)
Core average exit quality (fraction)
Core average void fraction (fraction)

2 030.000
61.600625.100
980.260
6 798 470.000
10 445.000
48 005.291
442.310113 560.700
2 871.24*
0.097
0.304

* Corrected for calibration and conversion errors

16

The initial water level above vessel zero (AVZ) was equal to 14.1478 m (557 in). This measured
level was the actual level inside the steam dryer shroud. The initial level AVZ was equal to 14.3256 m(564 in) for the narrow range measurement outside the steam dryer shroud. AVZ was the lowest
interior elevation of the vessel (bottom of lower plenum). Table 2.3-2 and Figure 2.3-1 provide theprocess computer P1 edit for the initial core relative axial power distribution to be used in Exercise 1.
Table 2.3-2. PB2 TT2 initial core average axial power distribution from P1 edit
Axial node number Axial location (cm)1017.62
2
122.863
138.104
153.345
168.586
183.82
7
199.068
114.309129.5410
144.7811
160.02
12
175.26
13
190.5014
205.7415
220.9816236.2217251.4618266.7019281.9420
297.1821
312.4222327.6623342.9024
358.14

Relative power
0.308051
0.616103
0.707754
0.773947
0.814681
0.880874
0.972526
1.066723
1.163467
1.260210
1.356953
1.407871
1.412963
1.402779
1.377320
1.328949
1.257664
1.188925
1.122733
1.031081
0.913971
0.763764
0.580460
0.290230

Figure 2.3-1. PB2 TT2 initial core axial average power distribution from P1 edit
1.6
1.4
Relative Power

1.2
1.0
0.8
0.6
0.4
0.2
0.0
0

5

10
15
Axial Nodes

17

20

25

2.4 Transient calculations
During the TT2 test, most of the important phenomena occur in the first five seconds of thetransient since it is a fast transient. For the purpose of the benchmark, the transient is required to be
simulated for a five-second time period. This approach simplifies the number of components requiredfor performing the analysis of TT2. The transient begins with the closure of the TSV. At some point
in time, the turbine BPV begins to open. The only boundary conditions imposed on the analysis should
be limited to the opening and closure of the above valves. Also, it should be noted that during thesimulation of Exercise 1, the set points of the SRVs are never reached. Since there is no neutron
kinetics model utilised in Exercise 1, the scram is not simulated explicitly. Rather, it is accounted for
implicitly in the provided power (or reactivity) time history to be used for this exercise. Table 2.4-1
shows the experimental time sequence of events during the transient.
Table 2.4-1. PB TT2 event timing (time in milliseconds)
TSV begins to close
TSV closed
Begins bypass opening
Bypass fully open
Turbine pressure initial response
Steam line A
Steam line D
Steam line pressure initial response
Steam line A
Steam line D
Vessel pressure initial responseCore exit pressure initial response

18

110
096
160
846
–
102
126
–
348
378
432
486

Chapter 3
METHODOLOGIES FOR COMPARATIVE ANALYSISAs mentioned in Chapter 1, each of the 14 participants submitted various data that were available
for statistical analysis. The submitted results were categorised by nature of the data. For Exercise 1,
the categories were: integral parameter values, 1-D distributions and time histories.
In addition, the submitted data could be distinguished according to the reference data used in thecomparative analysis. The reference data used in the Exercise 1 comparative analysis were called:
measured data, Exelon data and averaged data.
Measured data is the set of original recorded data during the second turbine trip transient test.
In the case of unavailable measured data, Exelon results (so-called Exelon data) were used as reference
results for the comparison. The main reason for this approach is that the Exelon results were alreadyextensively validated with the measured data. In other words, the Exelon data agree quite well with themeasured data and this agreement was published in the PECo Topical Report, which was submitted to
US NRC [7]. It should be noted that the current operator of PB2 NPP is Exelon Nuclear. The former
operator, the Philadelphia Electric Company (PECo), recently merged with Exelon Nuclear. In theevent of a lack of measured data and/or Exelon data for a given parameter, the reference solution foreach requested parameter is based on the statistical mean value (so-called averaged data) of all
submitted results.
3.1 Standard techniques for the comparison of results
In Exercise 1 of this benchmark, several types of data were analysed and the results of allparticipants were compared. The data types were:x Integral parameter values.
x One-dimensional (1-D) axial distributions.
x Time histories.
It was necessary to develop a suite of statistical methods for each of these data types, which wereapplied in the comparative analysis. What follows is a description of each of these methods.
3.1.1

Integral parameter values

These parameters include such values as core inlet enthalpy and core average pressure drop for
initial steady state conditions. Since measured data exist for both core inlet enthalpy and core averagepressure drop, standard deviation and figure of merit (FOM) are calculated in Eqs. (3.1-3.3).

19

¦ x x
V r
i

2

(3.1)

MeasuredN 1

where V is the standard deviation, xi is the data submitted by each participant and N is the total numberof received results.
The FOM is computed as follows:)i = ei/V(3.2)

ei = xi – xMeasured(3.3)

where ei is the deviation for each participant result.
3.1.2

One-dimensional (1-D) axial distributions

Exercise 1 compared only one 1-D axial distribution, which was the core average axial void fraction
distribution. The core average axial void fraction distribution is a function of height or number of axial
nodes, and it can be displayed as an x-y plot. Similar methods of statistical analysis described in theprevious section could be applied for each axial cell, the only difference being the reference data. In
this case and in the event of a lack of measured data, the Exelon data were taken as reference for thecomparison. Analyses were performed for each 1-D cell according to Eqs. (3.4-3.6):Vr¦ x xi

2(3.4)

ExelonN 1where xi is each participant’s data and N is the total number of received results. FOM is computed as:)i = ei/V(3.5)

ei = xi – xExelon(3.6)

For each participant, a table was prepared that shows the deviations from mean and FOM at eachaxial position.
3.1.3

Time histories

Six different sets of time histories were submitted by the participants for Exercise 1. The
submitted time histories included: dome pressure, core exit pressure, jet pump flow rate, core in-channel
flow rate, steam bypass flow rate and turbine inlet pressure. The reference data for dome pressurecomparison was the measured data. Since Exelon data were available for core exit pressure, jet pumpflow rate, core in-channel flow rate and steam bypass flow rate, the comparisons were based on the
Exelon data. It should be noted that for some of these categories there were measured data. However,the quality of this data (oscillating time histories) was not appropriate and the oscillatory behaviour ofthe data needed to be smoothed. The Exelon results show perfect agreement with the smoothedmeasured data. Due to the lack of measured and Exelon data, the statistical mean value of thesubmitted turbine pressure results (also called averaged data) was used as reference. The averaged data
were calculated according to the equation that follows.
20

N

x

¦

xi

(3.7)

i

N

where xi is each participant’s value for specified time interval and N is total number of received results.
3.2 ACAP analysis
The comparative analysis was performed for code-to-data and code-to-code comparisons using
the standard statistical methodology with the Automated Code Assessment Program (ACAP) [8].
ACAP is a tool developed to provide quantitative comparisons between nuclear reactor system code
results and experimental measurements. This software was developed under a contract with PSU andthe NRC for use in PSU’s code consolidation efforts. ACAP’s capabilities are described as follows:x Draws upon a mathematical toolkit to compare experimental data and NRS code simulations.
x Returns quantitative FOMs associated with individual and suite comparisons.
x Accommodates the multiple data types encountered in NRS environments.
x Incorporates experimental uncertainty in the assessment.
x Provides “event windowing” capability.
x

Accommodates inconsistencies between measured and computed independent variables(e.g. different time steps).

x

Provides a framework for automated, tuneable weighting of component measures in theconstruction of overall FOM accuracy.

ACAP is a PC and UNIX station-based application that can be run interactively on PCs with
Windows 95/98/NT, in batch mode on PCs as a Windows console application, or in batch mode on
UNIX stations as a command line executable.
The D’Auria Fast Fourier Transformation (FFT) and Mean Error (ME) methods were used in FOM
calculations for time histories [8,11]. Figure 3.2-1 shows a snapshot of the FOM configuration for
ACAP calculations in Exercise 1 of the benchmark. These methods are advanced techniques for analysisof time history data. Eqs. (3.8-3.13) represent the theory portion of the D’Auria FFT and ME methods.

21

Figure 3.2-1. FOM configuration in ACAPThe Discrete Fourier Transform (DFT) can be calculated as:

)m

1 N
) i e  2 SImi N
N i 1

¦

(3.8)

The D’Auria measures are as follows:x Average amplitude (AA)
M

AA

¦

P̂m  Ôm

m 0
M

¦ Ô

m 0

22

m

(3.9)

x Weighted frequency (WF)
(3.10)

M¦ P̂  Ô x f
m

WF

m

m

m 0

M

¦ P̂  Ô
m

m

m 0

where N is the number of data values, i is the sample index, )i is the value spaced 't apart, Oi is the ith
datum in the experimental set, Pi is the ith datum in the computed set, and fm is the frequency of mode m.
Mean error can be computed as:
ME

½
1 ­N
Oi  Pi ¾®
N ¯i 1¿¦(3.11)The D’Auria and ME FOM equations are outlined below:
FOM D’ Auria

1(3.12)

122·§ª§ K · º
¸
¨
2
AA


1
»
«
¸
¨
¸
¨
WF
¸
¨ ¬«
¹ ¼»
©
¹
©

where K is the constant used to weight the relative importance of the weighted frequency relative tothe average amplitude.
FOM ME

1
ME  1(3.13)

Please note that the statistical FOM given by Eqs. (3.2, 3.5) differs from the FOM calculatedusing Eqs. (3.12,3.13). FOM indicates that participant results are closer to the reference solution if theformer FOM is closer to zero, while the latter FOM must be closer to unity (1) to indicate better
agreement.
The following procedure was applied during the ACAP calculation:
Step 1 Data synchronisation was necessary due to the varying timesteps of submitted participantdata for the time histories. Regarding synchronisation, the cubic spline function was
written in Visual Basic for Applications (VBA) and a macro module was inserted into the
Excel workbook file [9]. Then, all participant results were set to time intervals of precisely
6 ms. The inserted module is outlined in the text that follows.

23

Function cubic_spline(input_column As Range, output_column As Range, _x As Range)
Dim input_count As Integer
Dim output_count As Integer
input_count = input_column.Rows.Countoutput_count = output_column.Rows.Count
If input_count <> output_count Then
000cubic_spline = "error"
000GoTo out
End If
ReDim xin(input_count) As Single
ReDim yin(input_count) As Single
Dim c As Integer
For c = 1 To input_countxin(c) = input_column(c)yin(c) = output_column(c)Next cDim n As Integer
Dim i, k As Integer
Dim p, qn, sig, un As Single
ReDim u(input_count - 1) As Single
ReDim yt(input_count) As Single
n = input_countyt(1) = 0u(1) = 0
For i = 2 To n - 1000sig = (xin(i) - xin(i - 1)) / (xin(i + 1) - xin(i - 1))000p = sig * yt(i - 1) + 2
000yt(i) = (sig - 1) / p
000u(i) = (yin(i + 1) - yin(i)) / (xin(i + 1) - xin(i)) - (yin(i) - yin(i - 1)) / (xin(i) - xin(i - 1))000u(i) = (6 * u(i) / (xin(i + 1) - xin(i - 1)) - sig * u(i - 1)) / p
000Next i
000qn = 0000un = 0000yt(n) = (un - qn * u(n - 1)) / (qn * yt(n - 1) + 1)
For k = n - 1 To 1 Step -1000yt(k) = yt(k) * yt(k + 1) + u(k)Next kDim klo, khi As Integer
Dim h, b, a As Single
klo = 1khi = nDok = khi - klo
If xin(k) > x Thenkhi = kElse
klo = kEnd Ifk = khi - klo
Loop While k > 1
h = xin(khi) - xin(klo)a = (xin(khi) - x) / hb = (x - xin(klo)) / h
y = a * yin(klo) + b * yin(khi) + ((a ^ 3 - a) * yt(klo) + (b ^ 3 - b) * yt(khi)) * (h ^ 2) / 6
cubic_spline = yout:
End Function

24

Step 2 In order to avoid the effects of differing participant initialisation on the comparative
analysis, actual values of the time history data were set to zero and they were called deltachanges. The delta changes were calculated by subtraction of initial value (at time zero)from all other values (as shown below) where t represents time.
Delta changes (t = 0.000) = value (t = 0.000) – value (t = 0.000)
Delta changes (t = 0.006) = value (t = 0.006) – value (t = 0.000)
Delta changes (t = 0.012) = value (t = 0.012) – value (t = 0.000)
Delta changes (t = 0.018) = value (t = 0.018) – value (t = 0.000)
Delta changes (t = 0.024) = value (t = 0.024) – value (t = 0.000)
Delta changes (t = 0.030) = value (t = 0.030) – value (t = 0.000)
Delta changes (t = 0.036) = value (t = 0.036) – value (t = 0.000)
Delta changes (t = 0.042) = value (t = 0.042) – value (t = 0.000)
Delta changes (t = 0.048) = value (t = 0.048) – value (t = 0.000)…

Step 3 The ACAP input file should be prepared properly and loaded to ACAP. Below is a
simplified example of an ACAP data file.
sync 1 this is synchronised data
title data MEASURED:
title data CEA:
title data FRA:
title data GRS:
title data IBE:
title data NET:
title data NFI:
title data NUP:
title data PSI:
title data PSU:
title data PSU/NRC:
title data TEP:
title data U.PISA:
title data UPV:
title data WES:
data
00.000000
0.006
0.005516
0.012
0.005516
0.018
0.000000…
data
00.000000
0.006
0.000000
0.012
0.000000
0.018
0.000001…
data
00.000000
0.006
0.000000
0.012
0.000000
0.018
0.000000…
…

Step 4 Using ACAP’s user friendly interactive options, a configuration must be selected and theFOM calculation should be performed in accordance with the reference data.
25

Chapter 4
RESULTS AND DISCUSSION OF EXERCISE 1

The tables and figures in this chapter provide a comparison of participant results with the reference
solution for the parameters that have the greatest effect on the initial steady state and transient scenario.
The tables show values of standard deviation and figure of merit (as defined in Chapter 3) for eachparticipant result and for a given parameter. This chapter contains figures that show the scatter of dataaround the reference solution.
Participant models used in Exercise 1 are presented in Table 4-1. This table is categorised byusage of bypass model, decay heat and power model as well as by the number of core channels in theparticipant system models. Core inlet enthalpy, core average pressure drop and core averaged axial
void fraction distribution are the selected parameters for steady state comparisons in this exercise. For
these parameters, Figures 4-1 through 4-6 graphically illustrate the agreement or disagreement ofparticipant predictions. Tables 4-2 through 4-5 show values, deviations and figures of merit for eachparticipant result and given parameter.
In the transient analysis, reactor vessel dome pressure, core exit pressure, jet pump flow rate (total
core flow rate), core in-channel flow rate, steam bypass flow rate (turbine bypass flow rate) and turbine
inlet pressure are the selected time histories for Exercise 1 comparisons. Time histories were analysed
for two different time intervals, which were 5 seconds and 1.5 seconds. Since this transient is a very fast
one, the study focused on the initial 1.5 s period after performing the analysis on the 5 s period.
Figures 4-7 through 4-40 compare the participant results with the reference solutions for each parameter.
Reference solutions are also compared with the averaged values of participant results. In addition,
Tables 4-6 through 4-17 show the ACAP FOM results for each parameter.
Although each of the 5 s time histories consisted of 833 data points and each of the 1.5 s time
histories consisted of 250 data points, few of these data points were marked in the figures. One marker
for each 50 points (for each 0.3 s time interval) was shown on the figures in order to make the figures
legible. For this purpose, a macro module was written in VBA for the Excel workbooks.
Participant marker types, marker frequencies and the thickness/colour of lines were unique andconsistent for all time history figures. For instance, the reference data (measured, Exelon or averaged). Part of the macro module is subsequently shown.
were always illustrated with the marker

27

Exercise 1 – Macro module
Sub Macro_Marker71()' Macro_Marker71 Macro' Macro recorded 10/2/2003 by rdfmg
ActiveChart.SeriesCollection(1).SelectActiveChart.SeriesCollection(1).Points(50).SelectWith Selection.Border.ColorIndex = 1.Weight = xlMedium.LineStyle = xlContinuous
End With
With Selection.MarkerBackgroundColorIndex = 1.MarkerForegroundColorIndex = 1.MarkerStyle = xlSquare.MarkerSize = 9.Shadow = False
End With
ActiveChart.SeriesCollection(1).SelectActiveChart.SeriesCollection(1).Points(100).SelectWith Selection.Border.ColorIndex = 1.Weight = xlMedium.LineStyle = xlContinuous
End With
With Selection.MarkerBackgroundColorIndex = 1.MarkerForegroundColorIndex = 1.MarkerStyle = xlSquare.MarkerSize = 9.Shadow = False
End With
ActiveChart.SeriesCollection(1).SelectActiveChart.SeriesCollection(1).Points(150).SelectWith Selection.Border.ColorIndex = 1.Weight = xlMedium.LineStyle = xlContinuous
End With
With Selection.MarkerBackgroundColorIndex = 1.MarkerForegroundColorIndex = 1.MarkerStyle = xlSquare.MarkerSize = 9.Shadow = False
End With
ActiveChart.SeriesCollection(1).SelectActiveChart.SeriesCollection(1).Points(200).SelectWith Selection.Border28

Exercise 1 – Macro module continued.ColorIndex = 1.Weight = xlMedium.LineStyle = xlContinuous
End With
With Selection.MarkerBackgroundColorIndex = 1.MarkerForegroundColorIndex = 1.MarkerStyle = xlSquare.MarkerSize = 9.Shadow = False***continues for all desired data points and for all other figures
illustrated on the Excel chart***End With
End Sub
Table 4-1. Participant models used in Exercise 1 of the BWR TT benchmark
Participants
CEA
Exelon
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

Bypass model
used?
Yes
Yes
Yes
Yes
Yes
Yes
Yes
Yes
Yes
Yes
Yes
Yes
Yes

Decay heat
used?
No
Yes
Yes
Yes
No
Yes
Yes
No
No
Yes
Yes
Yes
No

Power model
FissionFissionFissionFissionFissionFissionFissionFissionFissionFissionFissionFissionFissionNumber ofchannels
1
1
24
1
33
3
1
33
5
33
33
33
764

This chapter also contains the participants’ sequence of events for the turbine trip transient inTable 4-18. In addition, the time of maximum peaks for time histories are presented in Table 4-19 andthe maximum values of those peaks are given in Table 4-20.

29

Figure 4-1. Core inlet enthalpy
1212
Measured Value: 1209.06 Kj/kg

ENTHALPY (kJ/kg)

1210
1208
1206
1204
1202
1200
1198
1196

ES
W

PV
U

A
PI
S

P

PS
U

U

TE

/N

R

C

PS
U

I
PS

U
P
N

FI
N

ET
N

S
R
G

P
N
FA

E
EX

C

EA

1194

PARTICIPANTS

For core inlet enthalpy, key values are defined below.
Measured value:xMeasured = 1 209.06 kJ/kg

Standard deviation:¦ (x − xσ=±iMeasured

N −1

Figure of merit:)2= ±3.023

FOM = ei/σ where ei = xi – xMeasured

Table 4-2. Core inlet enthalpy – deviation (ei) and FOM

Value
ei (kJ/kg)
FOM

CEA
1 206.00-3.060-1.012

Exelon
1 209.06
0.000
0.000

Value
ei (kJ/kg)
FOM

PSI
1 208.24-0.820-0.271PSU
1 207.80-1.260-0.417FANP
1 208.30-0.760-0.251PSU/
NRC
1 207.70-1.360-0.450GRS
1 206.30-2.760-0.913

NETCORP
1 204.95-4.110-1.360

NFI
1 210.50
1.440
0.476

TEPSYS
1 200.80-8.260-2.732UPISA
1 208.50-0.560-0.185UPV
1 210.49
1.430
0.473

30

NUP
1 209.09
0.030
0.010
Westinghouse
1 206.27-2.790-0.923

Figure 4-2. Core pressure drop
0.160
Measured Value: 0.113561 MPa

PRESSURE DROP (M Pa)

0.140
0.120
0.100
0.080
0.060
0.040
0.020

ES
W

V
UP

IS

A

P
UP

TE

C
U/
NR

PS

PS

U

I
PS

P
NU

I
NF

T
NE

RS
G

NP

E
EX

FA

A
CE

0.000

PARTICIPANTS
For core pressure drop, key values are defined below.
Measured value:xMeasured = 0.113561 MPa

Standard deviation:¦ (x − xσ=±iMeasured

N −1

Figure of merit:)2= ±0.012

FOM = ei/σ where ei = xi – xMeasured

Table 4-3. Core pressure drop – deviation (ei) and FOM

Value
ei (MPa)
FOM

CEA
0.0960-0.018-1.500

Exelon0.1158
0.002
0.167

Value
ei (MPa)
FOM

PSI
0.1178
0.004
0.333

PSU
0.1130-0.001-0.083FANP
0.1130-0.001-0.083
PSU/
NRC
0.1197
0.006
0.500

GRS
0.12100.007
0.583

NETCORP
0.1406
0.027
2.250

NFI
0.13210.019
1.583

TEPSYS
0.1322
0.019
1.583

UPISA
0.1197
0.006
0.500

UPV
0.1159
0.002
0.167

31

NUP
0.1041-0.009-0.750Westinghouse
0.11490.001
0.083

Figure 4-3. Core average void fraction distribution
0.70

0.60

VOID FRACTION

0.50

0.40

EXELON
CEA

0.30

FANP
GRS

0.20

NETCORP
0.10

NFI
NUPEC

0.00
0

5

10

15

20

25

AXIAL NODES

Figure 4-4. Core average void fraction distribution
0.70

0.60

VOID FRACTION

0.50

EXELON

0.40

PSI
0.30

PSU
PSU/NRC

0.20

TEPSYS
U.PISA

0.10

UPV
0.00
0

5

10

15

AXIAL NODES

32

20

25

Figure 4-5. Void fraction distribution – averaged vs. Exelon data
0.7

0.6

VOID FRACTION

0.5

0.4

0.3
EXELONAVERAGE

0.2

0.1

0
0

5

10

15

20

25

AXIAL NODES

Figure 4-6. Void fraction distribution – Exelon and deviation
0.7

0.6

VOID FRACTION

0.5

0.4

0.3

0.2

0.1

0
0

5

10

15

AXIAL NODES

33

20

25

Table 4-4. Core average void fraction – Exelon and deviation
Axial nodes
1
2
3
4
5
6
7
8
9
10
11
12

Exelon
00000
00000
00000
0.00330.01710.03720.06490.10300.15620.20550.26610.3312

Deviation
2.92E-05
9.60E-04
4.41E-03
9.29E-03
1.66E-02
2.83E-02
4.02E-02
4.11E-02
4.46E-02
4.89E-02
4.88E-02
4.76E-02

Axial nodes
13
14
15
16
17
18
19
20
21
22
23
24

34

Exelon0.38180.4294
0.46860.50140.52900.55280.57280.58980.60470.61670.6261
0.6329Deviation
4.46E-02
4.04E-02
3.65E-02
3.28E-02
3.18E-02
3.01E-02
2.70E-02
2.47E-02
2.28E-02
2.25E-02
2.36E-02
2.82E-02

35

FANP

00.0000
00.0000
00.0000
00.861101.042201.219101.4030
01.771301.639001.627801.383200.712200.3004-0.2525-0.9890-1.4451-1.7484-2.2824-2.3185-2.4008-2.5219-2.4978-2.2288-2.7518

CEA

03.287700.040700.0047-0.3537-1.0294-1.3141-1.3184-1.3455-1.1031-0.7873-0.4939-0.5294-0.8700-0.6040-0.5644-0.5305-1.1321-0.5249-0.3259-1.0850-0.2939-0.0311-0.9364-0.0674

Nodes

1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24

00.0000
03.304203.272102.765301.671700.886600.4055
00.0535-0.5112-0.8793-1.1209-1.3634-1.4507-1.5668-1.4301-1.2378-0.9277-0.6478-0.3852-0.105300.153500.364400.483100.4255

GRS
00.0000
00.0000
00.0000-0.3552-0.9759-0.9435-0.6692-0.4282-0.208500.161600.243900.105000.0964-0.0322-0.1397-0.2439-0.3270-0.4252-0.5370-0.6397-0.7719-0.8444-0.8602-0.7801

NETCORP

0.04270.00040.03010.00430.29780.50420.60630.77470.67650.74420.65430.42420.4096
0.35570.36160.38480.38020.35680.35070.33080.25090.15330.01480.1826NFI
00.001900.000100.0000-0.2801-0.3718-0.026700.101300.1244-0.0563-0.0090-0.1566-0.4287-0.4666-0.5348-0.4941-0.3838-0.2252-0.1122-0.011200.0600
00.064600.0331-0.067900.1133

NUP
0.00000.00000.0000-0.35520-1.03010-1.31450-1.61440-0.12610
0.41800.93951.09580.97701.01190.94870.91610.89590.83100.76940.76550.75640.67910.55200.41170.1034PSI
10.0000
10.0008
10.0001-0.3552-1.0301-1.2324-0.9166-1.0093-1.1764-1.0746-1.2335-1.5126-1.6221-1.7921-1.9078-1.9873-1.9022-1.8962-1.9994-1.6461-1.7208-1.6845-1.6941-1.3474

PSU

Table 4-5. Core average void fraction FOM
PSU/
NRC
10.0000
10.0000
10.0000-0.3552-0.9561-1.1499-1.2818-1.6776-1.9536-1.8266-1.9168-2.0534-1.9377-1.8371-1.6197-1.3506-0.9792-0.6764-0.4089-0.114210.1667
10.4444
10.7381
10.8475
10.0000
10.0000
10.0680
10.1292-0.1687-0.1272
10.0572
10.1679
10.0785
10.1820
10.0963-0.1113-0.1054-0.1510-0.1425-0.1098-0.0535-0.0332-0.0222-0.0162-0.0614-0.1067-0.186410.0709

TEPSYS
0.00000.29150.4952
1.33491.3462
1.1091
1.0665
1.12770.96290.95850.81800.68350.45680.3066
0.21620.14520.35830.0354-0.01511-0.05251-0.12971
0.21260.15050.1537UPISA

10.4699
10.0548
10.2197
10.3308
10.1436
10.1892
10.2749
10.3450
10.1877
10.2251
10.0902-0.1654-0.2105-0.3139-0.3509-0.3673-0.3579-0.3882-0.4511-0.5212-0.6328-0.7069-0.7438-0.0748

UPV

Figure 4-7. Vessel dome pressure (5 seconds)
0.50
MEASURED

CEA

Delta Pressure Changes (MPa)

0.40
FANP

GRS

0.30
NET

NFI

0.20

NUP

0.10

0.00
0

1

2

3

4

5

time (s)Figure 4-8. Vessel dome pressure (5 seconds)
0.50
MEASURED
PSI

Delta Pressure Changes (MPa)

0.40

PSU
PSU/NRC

0.30

TEPSYS
U.PISA
UPV

0.20

WES

0.10

0.00
0

1

2

3
time (s)3645Figure 4-9. Vessel dome pressure – measured vs. averaged data (5 seconds)
0.50

Delta Pressure Changes (MPa)

0.40

0.30

MEASURED
0.20

AVERAGE

0.10

0.00
0

1

2

3

4

5

time (s)Table 4-6. Vessel dome pressure FOM – participant vs. measured data (5 seconds)
Participants
CEA
FANP
GRS
NETCORP
NFI
NUPEC
PSI
PSU
PSU /NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.828037
0.885103
0.852758
0.871570
0.881309
0.884493
0.842589
0.880791
0.866089
0.869565
0.857825
0.872835
0.877271

37

Mean error
0.982646
0.988965
0.987540
0.968806
0.966439
0.978238
0.912324
0.983431
0.975788
0.986657
0.976782
0.996351
0.971293

Figure 4-10. Vessel dome pressure (1.5 seconds)
0.50
MEASURED

CEA

Delta Pressure Changes (MPa)

0.40
FANP

GRS

0.30
NET

NFI

0.20
NUP

0.10

0.00
0.0

0.3

0.6

0.9

1.2

1.5

time (s)Figure 4-11. Vessel dome pressure (1.5 seconds)
0.50
MEASURED
PSI

Delta Pressure Changes (MPa)

0.40

PSU
PSU/NRC

0.30

TEPSYS
U.PISA
UPV

0.20

WES

0.10

0.00
0.0

0.3

0.6

0.9
time (s)

38

1.2

1.5

Figure 4-12. Vessel dome pressure – measured vs. averaged data (1.5 seconds)
0.50

MEASURED

Delta Pressure Changes (MPa)0.40

AVERAGE

0.30

0.20

0.10

0.00
0.0

0.3

0.6

0.9

1.2

1.5

time (s)Table 4-7. Vessel dome pressure FOM – participant vs. measured data (1.5 seconds)
Participants
CEA
FANP
GRS
NETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.756886
0.825679
0.782096
0.820257
0.850961
0.837275
0.819942
0.831920
0.882162
0.809503
0.824089
0.821020
0.860597

39

Mean error
0.928376
0.958118
0.940894
0.947105
0.972152
0.951302
0.934641
0.957261
0.9931000.980646
0.997984
0.966431
0.999617

Figure 4-13. Core exit pressure (5 seconds)
0.50

EXELON
CEA

Delta Pressure Changes (MPa)

0.40

FANP
GRS
NET

0.30

NFI
NUP
0.20

0.10

0.00
0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

4.5

5.0

time (s)Figure 4-14. Core exit pressure (5 seconds)
0.50

EXELON
PSI
PSU

Delta Pressure Changes (MPa)0.40

PSU/NRCTEPSYS
0.30

U.PISA
UPV
WES

0.20

0.10

0.00
0.0

0.5

1.0

1.5

2.0

2.5

3.0

time (s)

40

3.5

4.0

4.5

5.0

Figure 4-15. Core exit pressure – Exelon vs. averaged data (5 seconds)
0.50

Delta Pressure Changes (MPa)

0.40

0.30

EXELON
0.20

AVERAGE

0.10

0.00
0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

4.5

time (s)Table 4-8. Core exit pressure FOM – participant vs. Exelon data (5 seconds)
Participants
CEA
FANP
GRS
NETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.814507
0.860783
0.854806
0.890288
0.833606
0.828507
0.839428
0.854025
0.862058
0.824385
0.790233
0.881102
0.810725

41

Mean error
0.982300
0.998363
0.984252
0.977578
0.96538
0.961375
0.919174
0.988059
0.985494
0.976990
0.989411
0.987486
0.971534

5.0

Figure 4-16. Core exit pressure (1.5 seconds)
0.50

EXELON
CEA
FANP

Delta Pressure Changes (MPa)

0.40

GRS
NET
0.30

NFI
NUP
0.20

0.10

0.00
0.0

0.3

0.6

0.9

1.2

1.5

time (s)Figure 4-17. Core exit pressure (1.5 seconds)
0.50

EXELON
PSI
PSU

Delta Pressure Changes (MPa)0.40

PSU/NRCTEPSYS
U.PISA

0.30

UPV
WES
0.20

0.10

0.00
0.0

0.3

0.6

0.9

time (s)

42

1.2

1.5

Figure 4-18. Core exit pressure – Exelon vs. averaged data (1.5 seconds)
0.50

Delta Pressure Changes (MPa)0.40

EXELON
0.30

AVERAGE

0.20

0.10

0.00
0.0

0.3

0.6

0.9

1.2

time (s)Table 4-9. Core exit pressure FOM – participant vs. Exelon data (1.5 seconds)
Participants
CEA
FANP
GRS
NETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.811733
0.828060
0.802623
0.847156
0.815541
0.820173
0.835825
0.851292
0.859699
0.818468
0.798261
0.819631
0.791952

43

Mean error
0.961451
0.968435
0.942384
0.956179
0.976066
0.944192
0.938156
0.961372
0.999202
0.977226
0.998377
0.960493
0.969443

1.5

Figure 4-19. Jet pump flow rate (5 seconds)
2000

EXELON

CEA

1500

Delta Flow Rate Changes (kg/s)

FANP
1000

GRS

NFI

500

NUP
0

0

1

2

3

4

5

-500

-1000

-1500

time (s)Figure 4-20. Jet pump flow rate (5 seconds)
2000

EXELON
PSI
1500

Delta Flow Rate Changes (kg/s)

PSU
PSU/NRC
1000

TEPSYS
U.PISA
500

UPV
WES
0
0

1

2

3

-500

-1000

-1500

time (s)4445

Figure 4-21. Jet pump flow rate – Exelon vs. averaged data (5 seconds)
2000

EXELON

1500

Delta Flow Rate Changes (kg/s)

AVERAGE

1000

500

0

0

1

2

3

4

-500

-1000

-1500

time (s)Table 4-10. Jet pump flow rate FOM – participant vs. Exelon data (5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.632452
0.631828
0.643767
0.656687
0.559590
0.618890
0.623257
0.495413
0.598229
0.634360
0.621570
0.514555

45

Mean error
0.986841
0.962128
0.971611
0.991434
0.900471
0.939688
0.928375
0.863238
0.908011
0.987226
0.926363
0.8538355Figure 4-22. Jet pump flow rate (1.5 seconds)
2000

EXELON

CEA

1500

Delta Flow Rate Changes (kg/s)

FANP
1000

GRS

NFI

500

NUP
0

0.0

0.3

0.6

0.9

1.2

1.5

-500

-1000

-1500

time (s)Figure 4-23. Jet pump flow rate (1.5 seconds)
2000

EXELON
PSI
1500

Delta Flow Rate Changes (kg/s)

PSU
PSU/NRC
1000

TEPSYS
U.PISA
500

UPV
WES
0
0.0

0.3

0.6

0.9

-500

-1000

-1500

time (s)

46

1.2

1.5

Figure 4-24. Jet pump flow rate – Exelon vs. averaged data (5 seconds)
2000

EXELON
1500

Delta Flow Rate Changes (kg/s)

AVERAGE
1000

500

0

0.0

0.3

0.6

0.9

1.2

1.5

-500

-1000

-1500

time (s)Table 4-11. Jet pump flow rate FOM – participant vs. Exelon data (1.5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.632452
0.631828
0.643767
0.656687
0.559590
0.618890
0.623257
0.495413
0.598229
0.634360
0.621570
0.514555

47

Mean error
0.986841
0.962128
0.971611
0.991434
0.900471
0.939688
0.928375
0.863238
0.908011
0.987226
0.926363
0.853835

Figure 4-25. Core in-channel flow rate (5 seconds)
2000

EXELON

CEA

1500

Delta Flow Rate Changes (kg/s)

FANP
1000

GRS

NFI

500

NUP
0
0

1

2

3

4

5

-500

-1000

-1500

time (s)Figure 4-26. Core in-channel flow rate (5 seconds)
2000

EXELON
PSI
1500

Delta Flow Rate Changes (kg/s)

PSU
PSU/NRC

1000

TEPSYS
U.PISA

500

UPV
WES

0
0

1

2

3

-500

-1000

-1500

time (s)48

45

Figure 4-27. Core in-channel flow rate – Exelon vs. averaged data (5 seconds)
2000

EXELON
1500

Delta Flow Rate Changes (kg/s)

AVERAGE
1000

500

0
0

1

2

3

4

5

-500

-1000

-1500

time (s)Table 4-12. Core in-channel flow rate FOM – participant vs. Exelon data (5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.638910
0.649355
0.649255
0.656670
0.572426
0.643341
0.645504
0.522176
0.590639
0.652023
0.623379
0.537434

49

Mean error
0.991234
0.969817
0.969817
0.997138
0.915737
0.942567
0.961227
0.872593
0.904306
0.989703
0.936095
0.867945

Figure 4-28. Core in-channel flow rate (1.5 seconds)
2000

EXELON

CEA

1500

Delta Flow Rate Changes (kg/s)

FANP
1000

GRS

NFI

500

NUP
0
0.0

0.3

0.6

0.9

1.2

1.5

-500

-1000

-1500

time (s)Figure 4-29. Core in-channel flow rate (1.5 seconds)
2000

EXELON
PSI
1500

PSU

Delta Flow Rate Changes (kg/s)

PSU/NRC
1000

TEPSYS
U.PISA
500

UPV
WES

0
0.0

0.3

0.6

0.9

-500

-1000

-1500

time (s)

50

1.2

1.5

Figure 4-30. Core in-channel flow rate – Exelon vs. averaged data (1.5 seconds)
2000

EXELON
1500

Delta Flow Rate Changes (kg/s)

AVERAGE
1000

500

0
0.0

0.3

0.6

0.9

1.2

1.5

-500

-1000

-1500

time (s)Table 4-13. Core in-channel flow rate FOM – participant vs. Exelon data (1.5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.638910
0.649355
0.649255
0.656670
0.572426
0.643341
0.645504
0.522176
0.590639
0.652023
0.623379
0.537434

51

Mean error
0.991234
0.969817
0.969817
0.997138
0.915737
0.942567
0.961227
0.872593
0.904306
0.989703
0.936095
0.867945

Figure 4-31. Steam bypass flow rate (5 seconds)
800

EXELON

Delta Flow Rate Changes (kg/s)

CEA

FANP

600

GRS

NFI
400

NUP

200

0
0

1

2

3

4

5

time (s)Figure 4-32. Steam bypass flow rate (5 seconds)
800

EXELON
PSI

Delta Flow Rate Changes (kg/s)

PSU
600

PSU/NRCTEPSYS
U.PISA
UPV

400

WES

200

0
0

1

2

time (s)

3

52

4

5

Figure 4-33. Steam bypass flow rate – Exelon vs. averaged data (5 seconds)

Delta Flow Rate Changes (kg/s)

800

600

400

EXELON200

Average

0
0

1

2

3

4

time (s)Table 4-14. Steam bypass flow rate FOM – participant vs. Exelon data (5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.779968
0.885318
0.815970
0.824380
0.822632
0.759959
0.836952
0.789913
0.827886
0.819684
0.721839
0.793363

53

Mean error
0.865189
0.984254
0.937836
0.945642
0.911949
0.999926
0.968644
0.928899
0.967461
0.973376
0.998870
0.883658

5Figure 4-34. Steam bypass flow rate (1.5 seconds)
800

EXELON

Delta Flow Rate Changes (kg/s)

CEA

FANP

600

GRS

NFI
400

NUP

200

0
0.0

0.3

0.6

0.9

1.2

1.5

time (s)Figure 4-35. Steam bypass flow rate (1.5 seconds)
800

EXELON
PSI

Delta Flow Rate Changes (kg/s)

PSU
PSU/NRC

600

TEPSYS
U.PISA
UPV
400

WES

200

0
0.0

0.3

0.6

time (s)

0.9

54

1.2

1.5

Figure 4-36. Steam bypass flow rate – Exelon vs. averaged data (1.5 seconds)

Delta Flow Rate Changes (kg/s)

800

600

400

EXELON200

Average

0
0.0

0.3

0.6

0.9

1.2

1.5

time (s)Table 4-15. Steam bypass flow rate FOM – participant vs. Exelon data (1.5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.849663
0.9192410.864690
0.840617
0.845051
0.865657
0.880917
0.837732
0.8221720.857962
0.8172250.770214

55

Mean error
0.947216
0.999309
0.964516
0.951867
0.944996
0.963570
0.971740
0.973358
0.967200
0.966541
0.982058
0.852074

Figure 4-37. Turbine inlet pressure (5 seconds)
0.80

AVERAGE
CEA
0.60

Delta Pressure Changes (MPa)

FANP
GRS
0.40

NFI
NUP

0.20

0.00
0

1

2

3

4

5

-0.20

-0.40

time (s)Figure 4-38. Turbine inlet pressure (5 seconds)
0.80

AVERAGE
PSI
0.60

PSU

Delta Pressure Changes (MPa)

PSU/NRCTEPSYS
0.40

U.PISA
UPV
0.20

WES

0.00
0

1

2

3

-0.20

-0.40

time (s)56

45

Table 4-16. Turbine inlet pressure FOM – participant vs. averaged data (5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.768247
0.815086
0.781540
0.790493
0.765019
0.768060
0.731602
0.727749
0.752652
0.791697
0.752164
0.815964

Mean error
0.994379
0.922731
0.967382
0.989711
0.941782
0.892899
0.961210
0.968607
0.981105
0.971980
0.981043
0.999511

Table 4-17. Turbine inlet pressure FOM – participant vs. averaged data (1.5 seconds)
Participants
CEA
FANP
GRS
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

D’Auria FFT
0.774653
0.817505
0.780127
0.790156
0.764252
0.761436
0.710820
0.733067
0.756667
0.793776
0.755355
0.809388

57

Mean error
0.999697
0.927251
0.961526
0.995539
0.946661
0.887322
0.966465
0.974009
0.975201
0.977449
0.975139
0.994519

Figure 4-39. Turbine inlet pressure (1.5 seconds)
0.80

AVERAGE
CEA
0.60

Delta Pressure Changes (MPa)

FANP
GRS
0.40

NFI
NUP

0.20

0.00
0.0

0.3

0.6

0.9

1.2

1.5

-0.20

-0.40

time (s)Figure 4-40. Turbine inlet pressure (1.5 seconds)
0.80

AVERAGE
PSI
PSU

0.60

Delta Pressure Changes (MPa)

PSU/NRCTEPSYS
0.40

U.PISA
UPV
0.20

WES

0.00
0.0

0.3

0.6

0.9

-0.20

-0.40

time (s)

58

1.2

1.5

Table 4-18. Sequence of events
Participants
CEA
ExelonFANPGRSNETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse
Measured Data

TSVC
96
96
96
96
90
96
96
96
90
90
96
96
96
96
96

BVBO
60
60
60
60
60
60
60
60
60
60
60
60
60
60
60

Event (milliseconds)
BVFO
DPIR
846
423
846
432
846
400
846
420
850
516
846
403
846
510
852
444
846
420
846
438
852
372
846
398
846
398
846
342
846
432

CEPIR
360
486
390
354
342
406
396
450
408
396
372
390
390
438
486

Description of events: TSVC – turbine stop valve closed, BVBO – bypass valve begins to open, BVFO – bypass valve
fully open, DPIR – vessel dome pressure initial response (0.78% increase over initial value; % criterion based on reference
result), CEPIR – core exit pressure initial response (0.27% increase over initial value; % criterion based on reference result).

Table 4-19. Time of the maximum peakParticipants
CEA
ExelonFANPGRSNETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYS
UPISA
UPV

Dome
pressure
3.738
13.186*
3.000
2.718
2.890
3.027
3.999
2.976
3.366
3.144
3.442
2.736
2.900

Core exit
pressure
3.762
2.940
3.000
2.742
2.990
3.049
3.339
2.988
3.348
3.264
3.478
2.754
2.900

Time (seconds)
Jet pump
Core inlet
flow
flow
0.492
0.474
0.546
0.546
0.552
0.564
0.522
0.516
0.510
N/A
2.916
2.868
0.594
0.594
0.522
0.522
0.612
0.588
3.126
3.132
0.534
0.630
2.700
2.700
2.778
3.798

Bypass flow
1.176
0.864
0.954
1.278
N/A
1.002
0.846
1.410
1.272
0.672
1.128
0.894
0.882

Turbine inlet
pressure
1.212
N/A
3.318
2.238
N/A
0.114
3.816
1.392
0.198
2.598
3.108
1.302
2.346

* Since measured data is available, the measured dome pressure value was used as reference in lieu of the Exelon data.

59

Table 4-20. Maximum peak value (delta changes*)

Participants

Domepressure(MPa)

Core exit
pressure(MPa)

Jet pump
flow(Kg/s)

Core
in-channel
flow (Kg/s)

CEA
ExelonFANPGRSNETCORP
NFI
NUPEC
PSI
PSU
PSU/NRCTEPSYSUPISA
UPVWestinghouse

0.421
1110.448***
0.449
0.458
0.430
0.429
0.444
0.393
0.451
0.449
0.454
0.439
0.462
0.479

0.418
0.439
0.430
0.453
0.430
0.420
0.424
0.383
0.418
0.440
0.430
0.453
0.340
0.430

747.0
1250.31
519.3
514.0
700.0
554.2
1010.311666.11640.0
1086.41
705.0
534.3
400.6
1328.61765.000
1271.3071
470.836546.700
N/A
601.090941.333
1670.6551
739.000
1097.7191
664.137567.080455.0201302.8101

Steam
bypassflow(Kg/s)
673.8
647.7
648.9
609.6
N/A
595.6
566.3
605.2
650.5
577.3
600.1
649.4
635.2
466.7

Turbineinletpressure
(MPa)
0.500
N/A
0.569
0.555
N/A
0.564
0.568
0.553
0.531
0.574
0.512
0.626
0.552
0.626

MVCEP
(%)**
106.1
106.4
106.0
106.6
106.3
106.1
106.2
105.8
106.5
106.8
106.6
106.5
106.8
106.3

*

To avoid the effect of varying initialisation of participant data, all data are set to delta changes. The equation for this isdelta changes = data – initial value.
** MVCEP is the maximum value in per cent of the initial value of the core exit pressure.
*** Since measured data is available for dome pressure, measured dome pressure value was used instead of Exelon data.

The calculated results presented in this chapter describe the thermal-hydraulic system response ofparticipant models. A brief description of Exelon’s code, RETRAN, is provided below given that someof the Exelon code results were used as reference in the comparisons, especially for time histories.
In order to understand the comparisons, participant code descriptions (see Appendix A) must be taken
into consideration.
RETRAN, which stands for Reactor Transient System Analysis Computer Code, has undergone acomprehensive qualification program by the Electric Power Research Institute (EPRI) and its RETRAN
Working Group. Analyses of PB2 turbine trip transients were an important part of the RETRAN
qualification program.
Figure 4-41 shows the definition of core regions in the RETRAN skeleton input deck as well as
the flow paths of the reactor vessel, the reactor recirculation system and steam lines. A single volume(Volume 7) is used to model the steam space above the steam separators. The downcomer region isdivided into three volumes. The upper downcomer (Volume 8) models the region surrounding thesteam separators and includes the normal steam-water interface. The middle downcomer (Volume 9)
models the region surrounding the standpipes and represents the volume where feedwater and liquidflow from steam separator mixes. The lower downcomer models the region surrounding the core shroud
and jet pumps. Flow to recirculation loops and jet pump suctions originate from this volume.
A single volume (Volume 1) is used to model the fluid region below the core support plate (lower
plenum). The upper plenum region is above the upper guide plate and standpipes, which are modelled
as single volumes (Volumes 4, 5). Two-sided, passive heat conductors are used to model the materialof the shroud head and standpipes. A single volume is used to model the internal volume of the
211 steam separators. Since liquid flows from the separators, which include a steam carry-under, andenters the sub-cooled middle downcomer, the effects of carry-under can be ascertained without
introducing steam bubbles into the liquid downcomer.

60

Figure 4-41. RETRAN PB2 thermal-hydraulic model

The results, which are summarised in the figures and tables contained in this chapter, indicate theimportance of pressure transients for BWRs. The results also clearly establish the need for correct
system input models for related system codes. Statistical evaluation of the parameters contributed tothe quantitative comparative analysis.
The comparative analysis of PB2 TT2 provided detailed information on modelling challenges tosystem computer codes. In order to understand the comparisons and to perform meaningful analyses,
the parameters should be considered in conjunction with the information provided in Appendices A
and B. Appendix A provides descriptions of participant codes in terms of modelling capabilities(i.e. drift-flux models vs. six-equation or five-equation models, etc.). Appendix B provides participant
responses to a questionnaire, which clarify participant nodalisations and selected modelling options.
Table 4-1 provides general information about the participant models utilised in Exercise 1.
Although different numbers of channels were used by participants in their calculations, this should not
significantly affect Exercise 1 results. Even one channel would be sufficient for the simulation of thisexercise because the initial core average axial power profile and power time history during thetransient (obtained from experimental data) were used as input data in participant models.
61

Correct modelling of initial steady state is important – the TT2 test has been initiated from thesteady state conditions of the EOC 2. Participant results have shown excellent agreement for steadystate parameters. Although CEA, NETCORP and TEPSYS predictions seem to deviate more noticeably
from the reference results presented in Figures 4-1 and 4-2, the maximum deviation for the core inlet
enthalpy is less than 0.7% and less than 23% for the core pressure drop. The inlet enthalpy usuallydepends on the amount of steam carry-under from the separators. The transient response of the system
is very sensitive to separator modelling because this affects the initial location of the bulk boilingin core channels. The good agreement obtained in the core inlet enthalpy predictions is encouraging
for the participants to perform further analyses. The accuracy of the core pressure drop calculations
is affected by the predictions of the inlet enthalpy, which is related to the modelling of steam
separators/dryers and loss coefficients in the vessel and channels. Since the inlet enthalpy predictions
agree quite well, the observed discrepancies in the pressure drop predictions may come fromparticipant models other than the separator models.
The most challenging part of the steady state analysis is the prediction of the void fraction
distribution. This prediction depends on the modelling of core channels, vessel levels, steam dryers/
separators, jet pumps and recirculation loops, and also depends on the features of the participant
system codes. With all these challenges in mind, one can consider that the core average axial void
fraction distribution results agree very well. An interesting observation can be made regarding
Figure 4-5, which shows very good agreement of the average result over all participants’ predictions
as compared to the Exelon results. Among the void fraction results, only the FANP prediction exhibits
deviation from the reference result as shown in Figure 4-3. The reason may come from the calculationof the channel and vessel loss coefficients. Most of the results have shown slight differences at theaxially lower (bottom) part of the core, such as PSI’s prediction (Figure 4-4). Since there is noneutronic feedback for this exercise, the slight deviations may be explained by participants’ different
thermal-hydraulic flow and heat transfer models and especially sub-cooled boiling models.
In summary, steady state simulation of large systems is very difficult and requires extensive workon each component’s model. In addition, correct simulation is necessary of the control phenomena in
input models, such as water level control, recirculation pump speed, jet pump flows, feedwater flow
and steam line pressure.
The calculated transient results and the comparisons presented in this chapter not only provide an
opportunity to understand the phenomena behind the turbine transient tests but also to engage incomprehensive testing and examination of the capability of participant codes to simulate suchtransients. During the benchmark workshops, three important decisions were made by the participants
related to the comparative analysis of the time histories: 1) to analyse the time histories of deltachanges (instead of absolute parameter values) in order to avoid the impact of initial steady state
disagreements; 2) to provide two comparisons for 1.5 s and for 5 s; and 3) to provide two FOM estimates
based on two different methods (FFT and ME).
In general, it was demonstrated that the steam bypass line has a strong interaction with the main
steam line and the steam bypass line’s impact on the pressure responses depends on the relative positionof the TSV and BPV. It was observed that an acoustical pressure wave (oscillation) can exist in thereactor main steam piping due to the combined compressibility and momentum effects in the steam
line. The fundamental mode of the acoustical pressure oscillation is strongly excited as a result of turbine
trip, which causes a sharp initial pressure rise in the reactor as the first pressure wave enters the reactor
vessel. In addition, the fundamental acoustical mode propagates in the main steam piping, entering thereactor vessel and passing through the steam separators into the reactor core with relatively little
attenuation. All of these physical facts can be seen in participant results (Figures 4-7 through 4-40).

62During analysis of the results for the first exercise, it was obvious that the sources of deviations inparticipant predictions came from the differences in modelling the two key parameters – pressure
response and core flow response. Accurate prediction of the pressure response depends on the numberof models used by a given code. Models include: the steam line model (requires adequate nodalisationand treatment of momentum effects), the steam bypass system model and the steam separator model(requires proper treatment of steam separator inlet inertia and non-equilibrium effects at steam-water
surfaces). Prediction of core flow response is sensitive to the jet pump model used (requires proper
representation of the dynamic response of flow-to-pressure) and the core exit/separator region model(requires proper representation of the dynamic response of two-phase flow).
Figures 4-7 through 4-18 and Tables 4-6 through 4-9 present the comparative analysis of participant
vessel dome pressure and core exit pressure results with the reference data. In principle, the results arein good agreement. In particular, the good agreement in the times of occurrence of the initial riseindicates that the time effects (delays, rise times and frequencies) are modelled well. In addition, theamplitudes of the pressures are generally in good agreement. The discrepancies can be attributed to theshortcomings of the code models. For example, improving the slip model may increase the accuracy of
the results because the reactor pressure vessel may have a larger liquid inventory and lower steam
inventory, which tend to produce higher pressures.
If Tables 4-6 through 4-9 are analysed together, it can be concluded that some participant
predictions agree well with the reference solutions in the first 1.5 seconds of the transient while otherparticipant predictions agree well in the later part of the transient. The reason for this may be explainedin the following statements. The overall behaviour of the pressure is mostly affected by the transient
flow rate of steam through the turbine stop valves and bypass valves. Simulation of the opening andclosing of the valves and nodalisation of the steam lines play an important role in the calculations. The
discrepancies seen in the latter phases of these curves may arise from depressurisation rate mechanisms(e.g. critical flow models and steam formation models).
Participant results for jet pump flow rate (total core flow rate) and core in-channel flow rate are
generally consistent as shown in Figures 4-19 through 4-30 and in Tables 4-10 through 4-13. Although
the submitted results in the figures seem to have great deviations when compared with the referencedata, it should be noted that the illustrated results are based on delta changes (explained in Chapter 3).
If the initial values of the flow rates (such as 10 445 kg/s for total core rate) are considered, then the
maximum deviation is evaluated at less than 10%.
Discrepancies in flow rate responses can be analysed in 2 sections – deviations before 1.5 seconds
and after 1.5 seconds. Most problematic in the period before 1.5 seconds were predictions of peak times
and peak amplitudes as well as the simulation of wave propagation. The best agreement with reference
results in the flow rate prediction was achieved by the PSI results. As described in Appendix A, PSI
used RETRAN-3D, which is based on the code RETRAN (used by Exelon for reference calculations).
In the latter phases of the jet pump flow rate and core in-channel flow rate figures, the NUPEC,PSU/NRC and Westinghouse predictions were ~3-5% higher than the reference solution. The mostprobable reason for this disagreement is the input model (if the effect of the core bypass flow modelon the core flows is considered). In most models, the core bypass flow is directly related to the total
core flow. For instance, a decrease in core bypass flow causes an increase in total core flow rate.
Although steam bypass modelling is one of the most challenging parts of the exercise, thesubmitted results have shown that participant models can simulate steam bypass flow successfully.
The submitted results agree well as shown by Figures 4-31 through 4-36 and by Tables 4-14, 4-15.
The predictions shown in the figures deviated at the peak time and especially at the end of the transient.
63

Assumptions used in modelling the phenomena between the flow through steam bypass valve and thevalve opening area had a great effect on the results. In addition, the predictions were very importantbecause the bypass flow rate itself greatly influenced peak pressures.
Predictions of the participant models are also affected by the choice of bypass piping heat transfer
options [10]. As hot steam enters the bypass valves, significant heat is transferred to the cold piping
walls due to the large temperature gradient. This event causes slower pressurisation of bypass piping,which increases the affects of the critical flow through the bypass valve. The heat transfer to the pipe
walls causes an increase of the flow rate until walls reach the steam temperature. In most cases, pipes
are cold (i.e. their temperatures are close to the condenser temperature); however, there is evidence
that the piping walls remain at a high temperature for some time after steam bypass opening. Since theTT2 event followed TT1, during the opening of the bypass valves the heat transfer to the walls may be
negligible [7]. Finally, it was shown that the transient flow phenomena in the steam bypass system areof great importance.
As a parameter for the time histories, turbine inlet pressure predictions were compared with theaverage result. All results were in quite good agreement. Some slight deviations on the initial pressure
response were possibly a result of the differences in the action of TSV modelling. The modelling of
this action affects the prediction of certain time effects such as delays, rise times and frequencies.
The remarkable overall good agreement between reference and participant results for time histories
raised confidence in the participants’ system codes as well as in the developed system models. Moreover,the excellent agreement, which is illustrated in Figures 4-9, 4-12, 4-15, 4-18, 4-21, 4-24, 4-27, 4-30,
4-33, 4-36 and 4-39, between average and reference results not only indicated the success of thisbenchmark but also encouraged participants to take part in the remaining exercises.
In addition to time histories, sequence of events, maximum peak times for time histories andmaximum peak values can be found in Tables 4-18, 4-19 and 4-20. Although participant event timings
for TSVC, BVBO and BVFO were perfect as shown in Table 4-18, initial responses had slightdifferences. In parallel, time and maximum peaks presented in Tables 4-19 and 4-20 had discrepancies.
The possible reasons for these discrepancies were explained in the above paragraphs.
In conclusion, the results of the first exercise verified the participants’ system codes and systemmodels, and contributed to gaining a more in-depth knowledge of the capabilities of these codes.
Developing a more in-depth knowledge of coupled computer code systems is important because 3-D
kinetic/thermal-hydraulic codes will play a critical role in the future of nuclear analysis. Since thepurpose of this exercise was to initialise and test the thermal-hydraulic system model responses, thischapter proved that this exercise laid the foundation for conducting other benchmark exercises in aconsistent and systematic way.

64

Chapter 5
CONCLUSIONSIn order to meet the objectives for the validation of best-estimate coupled codes, a systematic
approach was introduced to evaluate the turbine trip transient. Such codes use separate temporal andspatial models as well as numerical methods for core neutronics, core thermal-hydraulics and system
thermal-hydraulic simulations. Therefore, the validation of the codes should include testing thesemodels for the defined transient for a given type of reactor (BWR in this case) as separate exercises ofthe overall benchmark.
The ultimate goal is to enable participants to initiate and verify these models before focusing on
the major objective, which is the testing of coupling methodologies in terms of numerics, temporal andspatial mesh overlays. This approach allows one to evaluate in a more consistent way the modelling of
combined effects (neutronics/thermal-hydraulics as well as core/plant interactions) by removing theuncertainties introduced with separate models.
In order to perform such a comprehensive validation of coupled codes, a multi-level methodology
is employed. This includes the application of three exercises, the evaluation of different steady states
and the simulation of different transient scenarios.
To summarise this work, Exercise 1 of the BWR TT benchmark was discussed in detail. The resultsof this benchmark problem were intended to assist in the understanding of the behaviour of next
generation computer codes. Overall, this benchmark has been well-accepted internationally, with
14 participants from 8 countries participating in the first exercise. The results submitted by participantsof Exercise 1 were used to make comparisons and a subsequent statistical analysis. This informationencompassed several types of data for thermal-hydraulic parameters at the initial steady state conditions
and throughout the turbine trip transient. These included: integral parameters, 1-D axial distributions
and time histories.
Due to the sensitive nature of this transient to small variations in initial parameters, the participants
were advised to follow the final specifications as closely as possible. In order to avoid possible
misinterpretations, the benchmark team provided further clarifications before the final results were
submitted. The motivation behind such clarifications was to narrow down as much as possible themodelling differences for the initial steady state conditions as well as for the transient scenario. This
strategy helped to obtain a cluster of solutions to be used as a basis for further comparative analysis.
A detailed assessment of the differences between calculated results submitted by the participants
for this exercise was presented in Chapter 4 of this report. Overall, the participant results for integral
parameters, core-averaged axial distributions and core-averaged time histories were in good agreementwith reference solutions, especially considering the approximations in some participant models, theuncertainties of some system parameters and the difficulties in interpreting some of the measured
responses. This is important from the fluid phenomena point of view because pressure and flow waves
play an important role in the early phase of the response to a turbine trip.

65

Since core inlet enthalpy is highly sensitive to core inlet temperature, the submitted results were
accepted as being in very good agreement. Besides, participant predictions for the core pressure drop
were quite close to the reference value if complexities of such system models were taken into account.
In the void fraction results, deviations were mostly observed in the lower (bottom) part of the core
because of differences in the participants’ sub-cooled boiling models. However, the average distribution
calculated from the participants’ submitted results almost matched the reference solution.
Generally, time history results and transient single values were in very good agreement with thereference data. Although some parameters seemed to have deviations, surprisingly, the average values
matched the reference solution well. We observed that more than 10 submitted sets of data could makethe comparative analysis meaningful in terms of the statistical approach.
In the time history comparisons, the deviations were mostly observed within the first 1.5 second
time period after the start of the transient. Since power was provided as a function of time (to eliminate
the impact of neutronic feedback effects), the deviations in the first 1.5 seconds may have come fromthe computer code control capabilities or from the approximations of participant models used in thistransient (especially simulating the bypass phenomena).
It was demonstrated that Exercise 1 fulfilled its objective, namely, to help the participants initialise
and test thermal-hydraulic system models to be used later in Exercise 3 (coupled 3-D neutron
kinetics/system thermal-hydraulic calculations of PB2 TT2). In addition, the participants utilised
Exercise 1 to perform a number of sensitivity studies on different modelling and numerical issues.
Such investigations were aimed toward evaluation of impact of the following effects and parameters:x

Thermal-hydraulic modelling issues: Turbine bypass line modelling; nodalisation of vessel,
steam line and steam separator region; TSV position and steam mass flow modelling;
thermal-hydraulic model – number of equations; void fraction model; steam separator inertia;
and jet pump modelling.

x

Thermal-hydraulic key parameters: Feedwater temperature; jet pump parameters; void
fraction in the bulk water (carry-under); core outlet pressure; active core pressure loss; core
inlet temperature; core inlet mass flow without bypass flow; sub-cooling; void generation rate;
and gas gap conductance.

xTimestep size: Fixed timestep of 6 ms versus using a variable timestep algorithm with amaximum timestep size imposed.

Exercise 1 provided the opportunity to study the impact of different thermal-hydraulic models oncode predictions and to identify the key parameters for modelling a TT transient. This in turn allowed
the evaluation of key parameters through the performance of sensitivity studies, which allowed theparticipants to develop a more in-depth knowledge of the capabilities of current generation best-estimate
thermal-hydraulic system codes.

66

REFERENCES[1]

Ivanov, K.N., T. Beam, A. Baratta, A. Irani, N. Trikouros, PWR MSLB Benchmark – Final
Specifications, Nuclear Energy Agency, NEA/NSC/DOC(97)15, October 1997.

[2]

Solis, J., K.N. Ivanov, B. Sarikaya, A. Olson, K. Hunt, Boiling Water Reactor Turbine Trip (TT)Benchmark Volume I: Final Specifications, Nuclear Energy Agency of the Organisation for
Co-operation and Development, NEA/NSC/DOC(2001)1, June 2001.

[3]“Transient and Stability Tests at Peach Bottom Atomic Power Station Unit 2 at End of Cycle 2”,
General Electric Company/Nuclear Energy Engineering Division, NP-564, Research Project1020-I, Topical Report, June 1978.

[4]

Larsen, N.H., “Core Design and Operating Data for Cycles 1 and 2 of Peach Bottom 2”, EPRI
NP-563, June 1978.

[5]

Carmichael, L.A., R.O. Niemi, “Transient and Stability Tests at Peach Bottom Atomic Power
Station Unit 2 at End of Cycle 2”, EPRI NP-564, June 1978.

[6]

Hornyik, K., J.A. Naser, “RETRAN Analysis of the Turbine Trip Tests at Peach Bottom Atomic
Power Station Unit 2 at the End of Cycle 2”, EPRI NP-1076-SR Special Report, April 1979.

[7]

Olson, A.M., “Methods for Performing BWR System Transient Analysis”, Philadelphia ElectricCompany, Topical Report PECo-FMS-0004-A (1988).

[8]

Kuntz, R.F., G.F. Kasmala, J.H. Mahaffy, “Automated Code Assessment Program: Technique
Selection and Mathematical Prescription”, Applied Research Laboratory, Pennsylvania State
University, Letter Report 3 to US NRC, April 1998.

[9]

Cubic_spline for Excel workbook, SRS1 Software, February 2003. Available at the website
http://www.srs1software.com.

[10] Akdeniz, B., K.N. Ivanov, “Analysis and Sensitivity Studies of Exercise 1 of OECD/NRC BWR
TT Benchmark”, TANSAO 87 (2002).
[11] Ambrosini, W., R. Bovalini, F. D’Auria, “Evaluation of Accuracy of Thermal-hydraulics CodeCalculations”, Energia Nucleare, Vol. 7, No. 2 (1990).

67

Appendix A
DESCRIPTION OF COMPUTER CODES USED FOR ANALYSIS
OF EXERCISE 1 OF THE NEA-NRC BWR TT BENCHMARK69

CATHARE (CEA, France)
The development of the CATHARE code was initiated in 1979 [1]. Code objectives include:•

Perform safety analyses with best-estimate calculations of thermal-hydraulic transients inpressurised water reactors.

•

Quantify the conservative analysis margin.

•

Investigate plant operating procedures and accident management.

•

Evaluate usage as plant analyser in full-scope training simulator providing real-time calculation.

The code is based on a two-fluid six-equation model. The presence of non-condensable gases canbe modelled by one to four additive transport equations. The code is able to model any kind ofexperimental facility or PWR, and is usable for other reactors (fusion reactor, BWRs, etc.)
The CATHARE code applications have been extended by coupling with other codes:•

Three-dimensional neutronics code – CRONOS [2].

•Core thermal-hydraulic code (component code) – FLICA [3].

•Severe accident code – ICARE2 [4].

CATHARE Main features
CATHARE has a modular structure. The modules are:•One-dimensional module to describe pipe flow.

•

Volume module, which is a two-node module.

•

Three-dimensional module to describe multi-dimensional effects in the vessel.

To complete the modelling of the circuits, sub-modules can be connected to the main modules(e.g. the multi-layer wall module), the point neutronics module, valves, safety valves, check valves,
flow limiters, boundary conditions, etc.
Physical description
All modules use the two-fluid model to describe steam-water flows and four non-condensable
gases that may be transported. Both thermal and mechanical non-equilibrium of the two phases aredescribed.
All kinds of two-phase flow patterns are modelled [5]. Heat transfer with wall structures and fuelrods are calculated, taking into account all heat transfer processes: natural and forced convection withliquid or gas, sub-cooled and saturated nucleate boiling, critical heat flux, film condensation for effects
of non-condensable gases, etc.
71

The range of parameters is rather large: pressure from 0.1-25 MPa, gas temperature from
20-2 000°C, fluid velocities up to supersonic conditions and duct hydraulic diameters from 0.01-0.75m.
System of equations
Mass, momentum and energy equations are established for any CATHARE module. They are
written for each phase. They are derived from exact local instantaneous equations, using some
simplification through physical assumptions as well as time and space averaging procedures. One tofour transport equations can be added when non-condensable gases are present.
Physical models are required to close the system of equations for CATHARE. Closure
relationships concern mass, momentum and energy exchanges between phases and between each phaseand the wall. As much as possible, physical closure laws are developed on an experimental basis.
Numerical scheme and solution procedure
The numerical method in the CATHARE code uses a first order finite volume/finite difference
scheme with a staggered mesh and donor cell principle. The time discretisation varies from the fully
implicit discretisation used in the 0-D and 1-D modules to the semi-implicit scheme used in the 3-D
module. Wall conduction is implicitly coupled to hydraulic calculations.
The non-linear system of equations is solved by a Newton-Raphson iterative method. In the lastversion of CATHARE (v1.5), the solution can be distributed over several processors in order to reduce
the CPU time by parallel computing.
References[1]

Houdayer, G., J.C Rousseau, B. Brun, “The CATHARE Code and its Qualification on
Analytical Experiments”, 10th Water Reactor Safety Research Information, Washington,
October 1982.

[2]

Lautard, J.J., S. Loubiere, C. Magnaud, “CRONOS: A Modular Computational System for
Neutronic Core Calculations”, IAEA meeting, Cadarache, France (1990).

[3]

Toumi, I., A. Bergeron, D. Gallo, E. Royer, D. Caruge, “FLICA4: A Three-dimensional
Two-phase Flow Computer Code with Advanced Numerical Methods for Nuclear Applications”,
Nuclear Engineering and Design, 200, 139-155 (2000).

[4]Guillard, V., F. Fhicot, P. Boudier, M. Parent, R. Roser, “ICARE/CATHARE Coupling:Three-dimensional Thermal-hydraulics of Severe LWR Accidents”, ICONE9, Nice Acropolis,
France (2001).

[5]

Bestion, D., “The Physical Closure Laws in the CATHARE Code”, Nuclear Engineering and
Design, Vol. 124, pp. 229-245 (1990).

72

S-RELAP5 (FANP, Germany)
S-RELAP5 is the standard tool of FRAMATOME ANP for the investigation of accident scenarios
in nuclear power plants, both of PWR and BWR. The range of applications covers LBLOCA on the
one end and reactor transients on the other end. S-RELAP5 includes all features of the RELAP5 code
family. It is a highly modular code where the user defines the granularity of the analysis by defining
an appropriate noding scheme.
The hydraulic part of such a noding scheme includes a concatenation of hydraulic volumes
connected by junctions. In the case of a two-phase state in the hydraulic volume, thermal
non-equilibrium is assumed. The two-phase flow between the volumes is non-homogeneous; thecorresponding phasical velocities are determined by solving momentum equations. This approachrequires the definition of local static pressure, with the pressure gradient being the source of
movement for each phase. Thermal properties as densities or internal energies are locally defined.
In S-RELAP5, flow maps and heat transfer modes are defined; they allow for the application of
constitutive relations as inter-phasical shear or heat transfer coefficients, which depend on them. A
highly sophisticated numerical solution scheme allows for the consideration of hydraulic systems withup to 500 hydraulic volumes. S-RELAP5 has a fully developed fuel model and a point kinetic model.
ATHLET (GRS, Germany)
The system code ATHLET is applied for analysis of behaviour of the whole plant under accident
conditions. It is a thermal fluid dynamic system code based on 1-D pipe components with a wide range
of applications. The applications are comprised of anticipated and abnormal plant transients, small andintermediate leaks as well as large breaks in PWRs and BWRs. The code offers the possibility ofchoosing between different models of fluid dynamics. The two-phase flow can be described with up toa full six-equation model for mass, energy and momentum of both phases, including models for
non-condensables. A boron tracking model for single and two-phase flow has been implemented. The
code structure is highly modular and allows for easy implementation of different physical models. The
basic modules include: thermo-fluid dynamics, neutron kinetics, general control simulation moduleand numerical integration method FEBE. ATHLET provides a modular network approach for the
representation of a thermal-hydraulic system. The possibility to implement connections among a groupof parallel channels in the reactor core allows for the modelling of cross flow inside the reactor core.
DNB/3D (NETCORP, USA)
DNB/3D is a FORTRAN 77 computer program that simulates the nuclear steam supply system of
a boiling water reactor under transient conditions. Geometry and system component options are provided
to represent any of the current BWR designs. The major features of DNB/3D are as follows:•

Point, 1-D and 3-D neutronic kinetics models including Doppler, moderator density andcontrol rod feedback along with standard decay heat models. Initially, sub-critical cores canbe represented. A power forced option is also provided.

•

Multi-node radial and axial coolant channels and fuel rods.

•

Fixed nodalisation in the reactor vessel and steam lines.

•

One-dimensional homogeneous equilibrium conservation of mass, energy and momentum.
73•Mechanistic non-equilibrium flow quality or profile fit model option to represent slip with thecoolant channels.

•

Steam line includes representation of the MSIV, turbine control and stop valves, RSV andsteam bypass valves.

•

Method of characteristics used to solve conservation equations in the steam lines.

•

Turbine representation.

•

Structural metal components represented. Reactor protection and safety systems represented.

•

Control systems represented. Complete self-initialisation of all components and models.

•

Provisions to represent a wide variety of transient and accident scenarios. Flexible restartcapability. British or Sl input/output option.

TRAC-BF1 (NFI, Japan)
The code used in the first exercise is TRAC-BF1/MOD1. TRAC-BF1/COS3D is not used in thisexercise but in the others. TRAC-BF1 is a best-estimate transient analysis code for analysing the fullrange of postulated accidents in boiling water reactor systems and related facilities.
TRAC-BF1 (NUPEC, Japan)
NUPEC used the TRAC-BF1 code in the first exercise. TRAC-BF1 is the latest public domain
BWR version of TRAC, which deals with thermal hydraulics, fuel heat transfer and plant system.
Thermal-hydraulics utilises the two-fluid model that solves six balance equations of mass, momentum
and energy for liquid and vapor phases. Two-phase flow in the core region is treated as 1-D parallel
vertical flows. The heat transfer model solves 1-D radial heat conduction equations. Heat transfer
coefficients at the cladding surface are subdivided into each flow regime. Material properties are basedon the code-installed MATPRO correlations. TRAC-BF1 has major BWR component modules: vessel,channel, separator/dryer, pump, pipe, valve, etc. The plant system can be composed of thesecomponent modules. The reactor pressure vessel is modelled with the vessel component in 3-D
cylindrical co-ordinates. Two numerical methods for thermal-hydraulic calculations are used. The firstis the standard finite differential method with staggered mesh, which is used for the space integrationof both fluid flow and heat conduction. The second is the time integration of fluid flow equations,which is performed by the semi-implicit scheme with the stability enhanced two-step (SETS) method.
RETRAN-3D (PSI, Switzerland)Within the STARS project at PSI, the code environment for the coupled 3-D reactor-kinetics/
thermal-hydraulics transient analyses of Swiss LWRs is based mainly on RETRAN-3D and
CORETRAN for both PWR and BWR systems. (Note that other codes are used for specificapplications, e.g. RAMONA for BWR stability.) Both codes play distinct roles in this environment –RETRAN-3D is used for analysis of coupled 3-D core/plant system transients while CORETRAN is74

used for core-only dynamic analysis. An important aspect is that both codes are based on an identicalneutronics algorithm, allowing usage of CORETRAN as an interface code to help prepare the 3-D coremodel for RETRAN-3D. This approach forms the basis of PSI 3-D transient analysis methodology.
Participation in the OECD/NRC Peach Bottom 2 Turbine Trip Benchmark was prompted by thefollowing considerations. First, the PSI methodology so far has only been assessed for neutronicallydriven transients and for a PWR system transient. Since the benchmark addresses a BWR transient
driven by system thermal-hydraulic perturbations, it extends the range of code assessment. Secondly,
the benchmark incorporates three different phases, which are well-suited to a comprehensive
assessment of all the participating codes, according to the PSI viewpoint. Consequently, PSI participatedin all three phases of the benchmark.
RETRAN-3D MOD003.1, which is used in phase 1 of the benchmark, was developed by EPRI to
perform licensing and best-estimate transient thermal-hydraulic analyses of light water reactors and ismaintained by CSA/USA. RETRAN-3D is used to analyse thermal-hydraulic transients and requires
numerical input data that completely describe the components and geometry of the system being
analysed. Input data include fluid volume sizes, initial flow, pump features, power generation, heatexchanger properties and material compositions. RETRAN-3D can calculate a steady state initialisationfrom a minimal amount of information. The steady state option computes volume enthalpies from asteady state energy balance with the restriction that only one enthalpy may be supplied per flow
system. The range of applications of RETRAN-3D also includes the spatial kinetics behaviour ofmulti-dimensional reactor cores. RETRAN-3D permits the analysis of systems with non-equilibrium
thermo-dynamic conditions and allows for the presence of non-condensable gases in the fluid stream.
TRAC-BF1/MOD1 (PSU, USA)
TRAC-BF1/MOD1: An advanced best-estimate computer program for boiling water reactor
accident analysis [1]The TRAC-BWR Code Development Program at the Idaho National Engineering Laboratory(INEL) is a program that develops versions of the Transient Reactor Analysis Code (TRAC) to
provide the US Nuclear Regulatory Commission and the public with best-estimate capability for the
analysis of accidents and transients in boiling water reactor systems and related experimental facilities.
This effort began in October 1979 and resulted in the first publicly released version of the code,
TRAC-BD1 v1-1 which was sent to the National Energy Software Center in February 1981. The missionof this first version of the code was to provide best-estimate capability for the analysis of design basis
loss-of-coolant accidents (LOCAs) in BWRs. The code provided a unified and consistent treatment of
the design basis LOCAs sequence beginning with the blowdown phase, through heatup, then refloodwith quenching and, finally, refill phase of the LOCAs scenario. New models developed for
TRAC-BD1 in order to accomplish its mission included: a) a full two-fluid, non-equilibrium,
non-homogeneous thermal-hydraulic model of two-phase flow in all parts of the BWR system,
including a 3-D treatment of the BWR pressure vessel; b) a detailed model of a BWR fuel bundle,
which includes other models (a radiation heat transfer model for thermal radiation between multiple
fuel rod groups, the inner surface of the channel wall and the liquid and steam phases in the channel;
a leakage path model; and a quench front-tracking capability for both falling films and bottom flooding
quench fronts on all rod groups and the inner surface of the channel wall); c) simplified models of
BWR hardware components such as the jet pump and separator/dryer; d) a counter current
flow-limiting model for BWR geometries; e) a non-homogeneous critical flow model; and f) flow
regime-dependent constitutive relations for the transfer of mass, energy and momentum between theliquid and steam phases in two-phase flow and between each phase and structure.

75

The mission of the second publicly released version of the code, TRAC-BD1/MOD1, wasexpanded to include not only large and small-break LOCAs but also operational transients andanticipated transients without scram (ATWS) to which point reactor kinetics is applicable. Models
developed to support the broadening of the mission scope for the TRAC-BWR codes included:•

Balance of plant models such as turbine, feedwater heaters and condenser.

•

A simple lumped parameter containment model.

•

Reactivity feedback model for use in the point reactor kinetics model.

•

Soluble boron transport model.

•

Non-condensable gas transport model.

•

Two-phase level tracking model.

•

Control systems model.

•

Generalised component-to-component heat and mass transfer models.

•

Improved constitutive models for the transfer of mass, energy and momentum between thetwo phases and between the phases and structure.

User convenience features, such as free-format input and extensive error checking of the input
data, were also included in TRAC-BD1/MOD1.
The mission of TRAC-BF1/MOD1 is the same as for TRAC-BD1/MOD1, but the newcapabilities built into this code make it more suitable for that mission. The new models andcapabilities in TRAC-BF1/MOD1 include:• Material courant limit violating numerical solution for all 1-D hydraulic components.
• Implicit steam separator/dryer model.
• Implicit turbine model.
• Improved interfacial heat transfer.
• Improved interfacial shear model.
• A condensation model for stratified vertical flow.
• One-dimensional neutron kinetics model.
• Improved control system logic and solution method.
In addition to these code improvements, a pre-load processor was written for TRAC-BF1, itsgraphic routines were improved for adaptation to the Nuclear Plant Analyzer (NPA) at INEL, andmore than 95% of the coding was converted to ANSI Standard FORTRAN 77.

76

TRAC-BF1/MOD1 can be applied to any BWR accident analysis or thermal-hydraulic test facility,
including those requiring reactivity feedback effects, control system simulation and/or balance of plantmodel. TRAC-BF1/MOD1 has been applied to experimental facilities from simple pipe blowdowns(e.g. Edwards pipe tests) to integral LOCA tests [e.g. two-loop test apparatus (TLTA) facility] andeven to multi-dimensional test facilities [e.g. slab core test facility (SCTF)]. BWR small-break LOCAs
have also been simulated with TRAC-BD1. The BWR fuel bundle model is quite versatile and has
been used to simulate not only BWR fuel bundles within a BWR reactor vessel but also stand-alone,
single-bundle experiments in which advanced bundle hydraulics and heat transfer models are required.
Reference[1]

Giles, M.M., G.A. Jayne, S.Z. Rouhani, R.W. Shumway, G.L. Singer, D.D. Taylor, W.L. Weaver,“TRAC-BF1/MOD1: An Advanced Best-estimate Computer Program Boiling Water Reactor
Accident Analysis Volume 1: Model Description”, NUREG/CR-4356 EGG-2626, pp. 1-1, 1-2,
May 1992.

TRAC-M (PSU/NRC, USA)
TRAC-M performs best-estimate analyses of loss-of-coolant accidents and other accident andoperational transients in pressurised light water reactors as well as neutronic thermal-hydraulicexperiments in reduced-scale facilities. Models used include multi-dimensional two-phase flow,
non-equilibrium thermo-dynamics, generalised heat transfer, reflood and reactor kinetics. Automatic
steady state and dump/restart capabilities are also provided.
The partial differential equations that describe two-phase flow and heat transfer are solved byfinite differences. The heat transfer equations are evaluated using a semi-implicit time-differencing
technique. The fluid dynamics equations in the spatial 1-D, 2-D and 3-D components use a multi-steptime-differencing procedure that allows the material courant limit condition to be exceeded. The finite
difference equations for hydrodynamic phenomena form a system of coupled, non-linear equations
that are solved by the Newton-Raphson iteration method. The resulting linearised equations are solved
by direct matrix inversion. For the 1-D network matrix, this is done via a direct full matrix solver. For
multiple vessel matrix, this is done via the capacitance matrix method using direct banded matrix solver.
The number of reactor components in the problem and the manner in which they are coupled isarbitrary. Reactor hydraulic components in TRAC-M include PIPEs, PLENUMs, PRIZERs(pressurisers), PUMPs, TEEs, VALVEs, and VESSELs with associated internals. HSTR (heatstructure) SLAB and ROD components are available to compute 2-D conduction and surface
convection heat transfer in Cartesian and cylindrical geometries. FILL and BREAK components areused to apply the desired coolant flow and pressure boundary conditions, respectively, for TRAC-M
steady state and transient calculations. There are also the SEPD (separator) and TURB (turbine,
TRAC-M/F77 only) components, which are to be replaced in the future TRAC-M/F90 development.
The current status of SEPD and TURB is indicated as appropriate in this document. Additional
component models that are under development are indicated later in this section.
The TRAC-M computer execution time is highly problem-dependent and is a function of the total
number of mesh cells, the maximum allowable timestep size and the rate of the neutronic
thermal-hydraulic phenomena being evaluated. For TRAC-PF1 and later versions, stability-enhancing
two-step numerics in 1-D hydraulic components allowed the material courant limit to be exceeded.
This enabled the usage of very large timesteps in slow transients when only 1-D hydraulic components
77are modelled. In TRAC-PF1/MOD2 and later versions, SETS numerics were also applied to themulti-dimensional VESSEL component to allow the material courant limit to be exceeded and verylarge timesteps to be used for all system models. This facilitates significant speedups of one or twoorders of magnitude for slow accident and operational transients when multi-dimensional VESSEL
components need to be modelled.
Reference[1]

Steinke, R.G., V. Martinez, N.M. Schnurr, J.W. Spore, J.V. Valdez, “TRAC-M/FORTRAN 90(Version 3.0) User’s Manual”, NUREG/CR-6722, pp. 2-1, 2-2, May 2001.

TRAC-BF1-ENTRÉE (TEPSYS, Japan)
TRAC-BF1-ENTRÉE is a parallel coupled system of the highly accurate 3-D neutron kinetic
code ENTRÉE and the versatile BWR simulator TRAC-BF1.
TRAC-BF1
The version we applied is basically the same as that used by other participants. The modifications
made by TEPSYS are:
1. Coupling interface with ENTRÉE.
2. Original direct heating model that works in co-operation with ENTRÉE.
3. Neutron kinetics integration timestep controlling algorithm that works in co-operation with
ENTRÉE.
4. Modifications in the control block INTM for modelling the SRV movement (Figure A-1).
ENTRÉEENTRÉE solves the two-energy group 3-D diffusion equation based on the transverse integrated
nodal expansion method. The higher order bundle flux is expanded either with the Legendre
polynomial functions or the Legendre semi-analytical functions. In this study, the non-linear iteration
method with Legendre semi-analytical functions will be applied because of its high level of accuracy.
Communication between two codes is realised by the PVM (parallel virtual machine) protocols, whichenable synchronisation of data sending and receiving in two processes.
Important numerical options together with plant valve parameters applied in this study aresummarised in Table A-1.

78

Table A-1. TEPSYS reference numerical options in ENTRÉE and valve parameters
Numerical options in ENTRÉE
Solution method
Non-linear iteration method
Neutron flux expansion function
Legendre semi-analytical functions
Transverse integrated flux
Quadratic polynomials
Assembly discontinuity factor
Yes
Decay heat model
ANS-1979
Moderator direct heating
Yes
Number of thermal-hydraulic regions
33 (in-channel) + 2 (bypass)
Cross-section library
NEA single table format (functionsof only instantaneous variables)
Valve parameters
Turbine stop valve
Closing delay time = 0 s
Closing stroke time = 96 ms
Turbine bypass valve
Opening delay time = 60 ms
Opening stroke time = 786 ms
Safety relief valves (pressure set points and Relief valve opening delay time = 400 ms
flow capacities specified in Ref. [4])Relief valve opening stroke time = 150 ms
Relief valve closing delay time = 0 ms
Relief valve closing stroke time = 0 ms
Safety valve opening delay time = 0 ms
Safety valve opening stroke time = 300 ms
Safety valve closing delay time = 0 ms
Safety valve closing stroke time = 0 ms

Figure A-1. TEPSYS modifications in control block INTM

79

RELAP5 (UPISA, Italy)
The light water reactor transient analysis code, RELAP5, was developed at the Idaho National
Engineering Laboratory for the US Nuclear Regulatory Commission. Code uses include: analyses
required to support rulemaking, licensing audit calculations, evaluation of accident mitigation strategies,
evaluation of operator guidelines and experiment planning analysis. RELAP5 has also been used as
the basis for a nuclear plant analyser. Specific applications have included simulations of transients in
LWR systems such as loss-of-coolant, anticipated transients without scram and operational transients(e.g. loss of feedwater, loss of offsite power, station blackout and turbine trip). RELAP5 is a highly
generic code that, in addition to calculating the behaviour of a reactor coolant system during atransient, can be used for simulation of a wide variety of hydraulic and thermal transients in both
nuclear and non-nuclear systems involving mixtures of steam, water, non-condensables and solutes.
The RELAP5/MOD3 code is based on a non-homogeneous and non-equilibrium model for thetwo-phase system and is solved by a fast, partially implicit numerical scheme to permit economical
calculation of system transients. The objective from the outset of the RELAP5 development effort was
to produce a code that included important first-order effects necessary for the accurate prediction ofsystem transients but that was also sufficiently simple/cost effective for parametric/sensitivity studies.
The code includes many generic component models from which general systems can be simulated.
The component models include pumps, valves, pipes, heat-releasing or absorbing structures, reactor
point kinetics, electric heaters, jet pumps, turbines, separators, accumulators and control systemcomponents. In addition, special process models are included for effects such as form loss, flow at anabrupt area change, branching, choked flow, boron tracking and non-condensable gas transport.
The system mathematical models are coupled into an efficient code structure. The code includes
extensive input checking capability to help the user discover input errors and inconsistencies. Also
included are free-format input, restart, renodalisation and variable output edit features. These userconveniences were developed in recognition that the major cost associated with the use of a systemtransient code is the engineering labour/time involved in accumulating system data and developing
system models, while the computer cost associated with the generation of final results is usually small.
The development of models and code versions that constitute RELAP5 has spanned ~17 years,
from the early stages of the RELAP5 numerical scheme development to the present. RELAP5
represents the aggregate accumulation of experience in modelling reactor core behaviour during
accidents, two-phase flow processes and LWR systems. Code development has benefited fromextensive application and comparison to experimental data in LOFT, PBF, Semiscale, ACRR, NRU
and other test programs. The latest version used in the Peach Bottom BWR TT Benchmark is
RELAP5/MOD3.3.
TRAC-BF1-VALKIN (UPV, Spain)
The coupled code is called TRAC-BF1-VALKIN and it can be used to simulate transients whileconsidering neutronic phenomena in 3-D geometry and thermal-hydraulic processes in multiple
channel 1-D geometry.
TRAC-BF1
As the thermal-hydraulic code we used the best-estimate code TRAC-BF1, which has a two-fluid,
six-equation model to account for the thermal-hydraulic phenomena.
80

To model the heat transfer in the fuel, an axial-radial heat transfer equation is utilised. The
thermal-hydraulics processes are modelled solving six balance equations of mass, momentum and energy
for liquid and vapour phases. For the numerical integration of the fluid flow equations, a semi-implicit
two-step method is used for time discretisation and a first order finite difference method withstaggered mesh is applied to discretise the spatial portion of the equations.
VALKINThe neutronic code used is the VALKIN module. This module has two options, MODKIN andNOKIN. MODKIN is based on a nodal modal method to integrate the time-dependent neutron
diffusion equation in the approximation of two energy groups. It is also based on the expansion of theneutronic flux in terms of the dominant lambda modes of a given configuration of the reactor core.
NOKIN option uses a one-step backward discretisation of the neutron diffusion equation.
NEM
Neutronic code NEM was used to compare the results obtained from the VALKIN code. NEM isa 3-D kinetic code based on the nodal expansion method.
POLCA-T (Westinghouse, Sweden)
POLCA-T is a coupled 3-D core neutron kinetics and system thermal-hydraulics computercode [1]. The code is able to perform steady state and transient analysis of BWR. To varying extents,
the code is based on models and tools for BWR and PWR analysis used in the POLCA7 [2], BISON
and RIGEL codes [3]. The code utilises new advanced methods and models in neutron kinetics,
thermal-hydraulics and numerics. The main features of the code can be summarised as follows:•

Full 3-D model of the reactor core. The neutronic models in the code are the same as those in
the well-known static core analyser POLCA7 with the addition of a proper 3-D kinetics model.

•

Advanced five-equation thermal-hydraulic model with a thermal non-equilibrium descriptionof the steam-water mixture and its coupling to the heat structures. Separate mass and energybalances of the phases. Drift flux model that can handle all flow regimes.

•

Heat transfer model that also works in post-dryout.

•

The gas phase consists of steam and non-condensable gases. The liquid phase contains
dissolved non-condensable gas.

•

Boron transport model.

•

Use of the same thermal-hydraulics model for core and plant systems.

•

Two-dimensional fuel rod heat transfer model, gas gap model consistent with design code andcomplete range of heat transfer regimes. The code is able to model all plant heat structures.

•

Dryout and DNB correlations (using pin power distributions model).

81•Full geometrical flexibility of the code: volume cells, flow paths, heat structures, materials,
pumps, measurements, controls, etc. are all input data. Code is able to analyse different power
plants and test facilities.

•

Balance of plant, control and safety systems: reactor pressure vessel, external pump loops,
steam system, feedwater system, ECC systems and steam relief system are modelled todesired detail. A large and valuable set of existent input models can be used.

•

Stable numerical method that allows long timesteps, which is also used for a steady statesolution. Low dependence on the size of the timestep since the implicit numerical integration
is close to second order by means of θ-weighting.

Due to the above mentioned features, the POLCA-T code makes possible a comprehensive
approach to plant analysis with full consistency in steady state and transient calculations. Consistency
in BWR core and system modelling, when transient analyses are performed, is achieved by using thesame basic model and the same design database for core data (cross-section data, burn-up, xenon andother 3-D distributions obtained from depletion calculations), fuel thermal-mechanical behaviour(properties) and system data. Thus, full consistency is also possible between predicted core and systemparameters and their behaviour in a very wide range of phenomena and processes important both fordesign and safety analyses.
Application areas of the POLCA-T code consist of three groups: 1) BWR steady state core design,
2) BWR stability, transients and accident analyses and 3) modelling of experimental test facilities.
Applications for BWR steady state core design are covered by POLCA7’s broad capabilities:•

Evaluation of reactivity and power distribution at cold and hot core conditions.

•

Detailed thermal-hydraulic analysis.

•

Control operations (reactivity search modes including boron, power, flow, control rod, axial
offset and minimum boron control).

•Detector simulation.

•

Evaluation of fuel pins, pellet powers and burn-up.

•

Evaluation of peaking factors based on pin results.

•

Dryout and DNB margin calculations (based on pin power distributions).

•

Fission heat load parameters (margins) calculations.

•

Pellet cladding interaction calculations.

•

Depletion calculation with tracking of most important fissile isotopes and fission products.

•

Fuel bundle, control rod, fuel channel and fixed in-core detector depletion.

•

Shutdown margin evaluation.

82

•

Shutdown cooling.

•

Xenon transients.

•

Reactivity coefficients (void, burn-up and moderator temperature).

Applications for BWR safety analysis include operational transient, stability, RIA, ATWS,ATWC and LOCA as follows:•

Feedwater flow increase/temperature decrease transients.

•

Loss of feedwater flow transients.

•

Pressure increase transients – analysis on cladding.

•Pressure increase transients – analysis on the reactor coolant pressure boundary.

•Pressure decrease transients.

•

Recirculation flow increase transients.

•

Recirculation flow decrease transients.

•

Control rod withdrawal error.

•

Inadvertent loading transients.

•

Control rod drop accident.

•

Stability analysis.

•

Anticipated transients without scram.

•

Anticipated transients without control rods (ATWC).

The POLCA-T code is well-adapted to analyse the type of scenarios with a number of failing
control rods. The boron transport model makes it possible to analyse different types of boron shutdown
scenarios. Applications for the modelling of separate test facilities (e.g. FRIGG) are foreseen not only
for code validation but also for pre-test analysis and experiment planning and optimisation.
References[1]

Panayotov, D., U. Bredolt, P. Jerfsten, “POLCA-T – Consistent BWR Core and Systems
Modelling”, ANS/AESJ/ENS Int. Conf. Top Fuel 2003, Paper 410, Wurzburg, Germany,
16-19 March 2003.

[2]

Lindahl, S-Ö., E. Müller, “Status of ABB Atom’s Core Simulator POLCA”, Int. Conf.
PHYSOR96, Mito, Japan, 16-20 September 1996.

83

[3]

Wijkström, H., “ABB Atom’s New Code for 3D Static and Transient Analysis”, Proceedings ofthe German Nuclear Society Workshop on Thermal and Fluid Dynamics, Reactor Physics andComputing Methods, Rossendorf, Germany, 31 January through 1 February 2000.

[4]

Panayotov, D., “OECD/NRC BWR Turbine Trip Benchmark: Simulation by POLCA-T Code”,
PHYSOR2002 International Conference on the New Frontiers of Nuclear Technology: Reactor
Physics, Safety and High-Performance Computing, Track H-2, Paper 3C-02, Seoul, Korea.

84

Appendix B
QUESTIONNAIRE FOR EXERCISE 1 OF
THE NEA-NRC BWR TT BENCHMARK85

QUESTIONNAIRE FOR EXERCISE 1

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)c) Please provide in detail the nodalisation of the steam line.

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
b) Which neutronic/power initial and transient boundary conditions are used and how?

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
b) How are the jet pumps modelled?
c) How is the steam separator modelled?
d) How are the turbine stop valve closing and steam bypass system modelled?
e) What were the difficulties encountered during the component design modelling?

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
c) Specific features of the codes used?
d) Number of solutions submitted per participant and how they differ?
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

87

CATHARE2 v1.5 (CEA, France)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
Reactor pressure vessel is modelled by 1-D and 0-D connected modules. Core is represented withone 1-D channel using 24 nodes.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
Big components like plenums and condenser are modelled by 0-D modules. The core, standpipes,
downcomer, recirculation loops and steam lines are represented with 1-D modules. Junctions
enable to connect each 1-D and 0-D component.
Figure B-1. CEA nodalisation diagram

88c) Please provide in detail the nodalisation of the steam line.
All parameters such as diameters, lengths and volumes are based on RETRAN skeleton deck of
BWR TT specifications [NEA/NSC/DOC(2001)1, Appendix A, Skeleton Input Deck].
The steam line between RPV and TSV is meshed with 200 elements (length of each element~1 meter). The steam line between TBV and condenser has 100 elements.
2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
A steady state condition is achieved by controlling the mass flow rate in the core and in therecirculation loops as well as by controlling the pressure in the steam dome.
The boundary conditions are:•Feedwater flow rate.

•

Pressure downstream to the TSV.

•Core power (axial distribution).

Transient boundary conditions are:•Evolution of feedwater flow.

•

Evolution of power.

•

Loss coefficient for TSV and BPV in order to respect closing and opening dynamics given
by BWR TT benchmark specifications.

b) Which neutronic/power initial and transient boundary conditions are used and how?
Power initial and transient boundary conditions given in BWR TT specifications are used (i.e. see
file nfpower_exercise1).
3.

Component design and modellinga) How many core bypass channels are used in this exercise?
One core bypass channel is used for Exercise 1.
b) How are the jet pumps modelled?
Jet pumps are modelled by 1-D and TE modules, taking into account momentum source.

89c) How is the steam separator modelled?
The steam separator model utilised is the CATHARE simple separator model with constantcarry-under ratios.
d) How are the turbine stop valve closing and steam bypass system modelled?
TSV and BPV are modelled by a head-loss coefficient, which is adjusted to simulate the valve
closing and opening.
e) What were the difficulties encountered during the component design modelling?
No particular difficulty was encountered for global modelling.
4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
The maximum timestep value for the transient calculation is 10–4 s. A sensitivity study showedthat results were not modified if the timestep was kept lower than 0.005 s.
c) Specific features of the codes used?
CATHARE is a best-estimate code for LWR transient analysis. The code is based on a six-equation,
two-phase flow model with a set of closure laws assessed on a large number of experiments.
d) Number of solutions submitted per participant and how they differ?
One solution for Exercise 1.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

90

S-RELAP5 (FANP, Germany)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
See the nodalisation schema below (Figure B-2).
Figure B-2. FANP nodalisation diagram

91b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
Close to RETRAN nodalisation schema.
c) Please provide in detail the nodalisation of the steam line.
See the FANP nodalisation diagram (Figure B-2).
2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Boundary conditions provided by the benchmark team.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Boundary conditions provided by the benchmark team.

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
One.
b) How are the jet pumps modelled?
S-RELAP5 standard component.
c) How is the steam separator modelled?
S-RELAP5 standard component.
d) How are the turbine stop valve closing and steam bypass system modelled?
Turbine stop valve – linear closure within 96 ms.
Steam bypass system – modelled in such a way that the steam outflow value of RETRAN(provided by the benchmark team) matched the final specifications.
e) What were the difficulties encountered during the component design modelling?

4.

Generala) Deviations from the updated final specifications?
None.

92

b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
Six milliseconds (6 ms).
c) Specific features of the codes used?
d) Number of solutions submitted per participant and how they differ?
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

93

ATHLET (GRS, Germany)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
One-dimensional (1-D) nodalisation, two T-H channels in the core.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
See Figures B-3 and B-4 for the nodalisation schemas of the pressure vessel and steam line.
c) Please provide in detail the nodalisation of the steam line.
Figure B-3. GRS reactor pressure vessel nodalisation schemaDRYER

SEPARATOR
D.COMER1

D.COMER2

U.PLENUM4
F-FEEDWAT
U.PLENUM3
U.PLENUM2

D.COMER3
JETP-IJET-P-INOZZLE1NOZZLE2

U.PLENUM1REFBYPASS
CORECHAN

D.COMER4
DRI.LINE1

DRI.LINE2
D.COMER6

D.COMER5

D.COMER7LOOP1BYPASSPLEN
COREINBYCOREINL.PLENUM2
CON.R.PIPE

LOOP2

DIFFUSOR-I

L.PLENUM1

94

Figure B-4. GRS steam line nodalisation schema

DRYER

B-TDV
MSL
MTV

2.

BYPLINE

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Mass flow rate at core inlet = 10 580.89 kg/s.
Bypass mass flow rate = 841.8 kg/s.
Pressure at core dome – as measured at steady state.
Core pressure losses adapted to 0.82 bar and mean void to 30.4%.
Core inlet enthalpy – as specified.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Axial power distribution – as specified.
Power history – as specified.

3.

Component designa) How many core bypass channels are used in this exercise?
One bypass channel.

95b) How are the jet pumps modelled?
The model consists of two symmetrical recirculation loops. The mass flow rate through the jetpumps and the pressure difference are adjusted to specifications (see nodalisation schema).
c) How is the steam separator modelled?
The steam separator is modelled by a corresponding module of ATHLET.
d) How are the turbine stop event and steam bypass flow simulated?
The turbine stop valve closing is modelled by a linear change of the flow area within the closure
time. The steam bypass valve is opened by a linear change of the flow area within the opening
time. It is assumed that critical flow conditions constantly exist and that in fully opened condition
the mass flow rate is ~600 kg/s.
e) What were the difficulties encountered during the component design?
Great effort was made to approximate the pressure distribution within the reactor vesselcorresponding to the given values from the plant.
4.

Generala) Deviations from the updated final specifications?
No considerable deviations.
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
c) Specific features of the codes used?
Five-equation model of the thermal-hydraulics.
d) Number of solutions submitted per participant and how they differ?
Only one solution delivered.
f) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

96

DNB/3D (NETCORP, USA)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the plant
General model illustrated in Figure B-6.
a) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
One-dimensional (1-D) homogeneous equilibrium.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
There are six volumes in the reactor vessel and 24 vertically stacked volumes in the core that
represent the core average fuel channel. See Figure B-5.
c) Please provide in detail the nodalisation of the steam line.
See Figure B-7. Note that there are seven piping volumes and each piping volume is subdivided(using mesh points for solving the conservation equations with the method of characteristics) into
sub-volumes of ~10-15 feet in length.

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
The thermal-hydraulic boundary conditions that were used are: 1) feedwater flow (see Table 5.3.3
of final specifications) and 2) steam bypass flow (from the BWR TT ftp site).
b) Which neutronic/power initial and transient boundary conditions are used and how?
The total core power boundary condition was used (file provided on the BWR TT ftp site).

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
One – represents the core average bypass channel (see Figure B-5).
b) How are the jet pumps modelled?
The jet pump representation is shown in Figures B-8, B-9 and B-10.
c) How is the steam separator modelled?
The steam separator representation is shown in Figures B-8, B-9 and B-10.

97d) How are the turbine stop valve closing and steam bypass system modelled?
The turbine stop valve position closing was assumed to close linearly with time; however, thenon-linear area as a function of position was included in the model. The steam bypass flow was
based on the boundary condition as noted in Figure B-10.
e) What were the difficulties encountered during the component design modelling?
No difficulties were encountered in the component modelling.
4.

Generala) Deviations from the updated final specifications.
There were no known deviations from the final specifications.
b) User assumptions. Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
The thermal-hydraulic timestep size was 0.01s for reactor vessel regions and 0.001 s for steamline regions.
c) Specific features of the codes used?
See the code description for DNB/3D in Appendix A.
d) Number of solutions submitted per participant and how they differ?
Only one solution was submitted.
e) If your results are significantly different (> 3σ) than the measurement or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
There are no known significant differences.

98

Figure B-5. NETCORP reactor vessel nodalisation

Figure B-6. NETCORP schematic of BWR NSSS representation

99

Figure B-7. NETCORP’s method of characteristics of main system

100

Figure B-8. NETCORP schematic of reactor vessel geometry for jet pumps

101Figure B-9. NETCORP flow diagram within reactor vessel for jet pumps

Figure B-10. NETCORP DNB/3D equations

102

TRAC-BF1 (NFI, Japan)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
RPV is modelled with 10 levels for vertical direction and two rings for radial direction using the
VESSEL component of TRAC-BF1. The T-H core is modelled with 33 channels using CHAN
components. Every CHAN component has 26 nodes in total – 24 nodes for active fuel, two nodes
for lower and upper tie plates.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
The nodalisation lines (in orange) of Figure B-11 indicate the nodalisation of RPV.
Top of level 1: Bottom of guide tube
Top of level 2: Bottom of downcomer
Top of level 3: Bottom of reactor core
Top of level 4: Top of jet pump
Top of level 5: Top of reactor core

Top of level 6: Top of shroud headTop of level 7: Bottom of separator
Top of level 9: Top of dryer
Top of level 10: Top of RPV

Figure B-11. NFI TRAC-BF1 nodalisation diagram
Nodalization

SEPD

TEE, VALVE, BREAK

Main SteamPIPE, FILL

Feed Water

CHAN#1-#33
JETP

PUMP

VESSELTRAC-BF1
103c) Please provide in detail the nodalisation of the steam line.
There are seven nodes between RPV and TSV, and 11 nodes between RPV and the condenser.
All parameters such as diameters, lengths and volumes are based on the RETRAN skeleton deckof the BWR TT specifications [NEA/NSC/DOC(2001)1, Appendix A, Skeleton Input Deck].
Figure B-12. NFI TRAC-BF1 steam line nodalisation

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Steady state condition is achieved using three element controls, namely, pressure control,
feedwater control and recirculation control, which are in agreement with the initial conditionindicated by the BWR TT specifications. Therefore, initial conditions of core exit pressure, waterlevel and total jet pump flow are in good agreement with the specifications. As for core inlet flow
and bypass flow, each flow rate and core inlet enthalpy are calculated by TRAC-BF1
simultaneously based on core power distribution, leak flow, two-phase friction pressure drop andpressure drop of SEO at the central and peripheral channels.
Transient boundary conditions such as pressure increase are caused by the closing of TSV and theopening of BPV, which cause pressure propagation to increase through the main steam line andthe bypass line. As a result, core exit pressure changes but core total power is given as input data.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Neutronic/power initial and transient boundary conditions given as BWR TT specifications areused (i.e. see file nfpower_exercise1).

104

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
Core bypass channels are not used as CHAN but as VESSEL components. Levels 3 and 4 of
VESSEL correspond to the bypass region. We neglect direct heating in core bypass region.
b) How are the jet pumps modelled?
The JETP component is used. The specification is the same as Table 3.1.2.2 and Figure 3.1.2.1 of
the BWR TT specifications. The junction loss coefficients (jet pump nozzle, suction, etc.) areutilised using the TRAC-BF1 recommended value.
c) How is the steam separator modelled?
The SEPD component is used. The simple separator model in TRAC-BF1 is selected, so thecarry-under ratio keeps constant during the transient.
d) How are the turbine stop valve closing and steam bypass system modelled?
The characteristics of Tables 5.2.1 and 5.2.2 of BWR TT benchmark specifications are used for
TSV and BPV, respectively.
e) What were the difficulties encountered during the component design modelling?
As for Exercise 1, there is no outstanding difficulty for component design modelling.

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
The minimum and maximum timestep value of TRAC-BF1 is 3.0 × 10–5s and 3.0 × 10–1 s,respectively. The actual timestep value is changing between them during the calculation.
c) Specific features of the codes used?
TRAC-BF1 is a best-estimate code for BWR transient or accident analysis. This code includes
a full non-homogeneous, non-equilibrium, two-fluid thermal-hydraulic model of two-phase flow.
d) Number of solutions submitted per participant and how they differ?
It is our view that the difference between the participants is caused by two main reasons. One isthe difference between codes and their models. The codes would be different in physical models
versus thermal-hydraulic models. Another is the difference of nodalisation. The modelling of
main steam lines/RPV or pressure loss profile along pipelines affects pressure propagation/increase.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
105

TRAC-BF1 (NUPEC, Japan)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
The reactor pressure vessel is modelled in 3-D cylindrical co-ordinates with two radial rings and
15 axial levels. Thermal-hydraulic channels were treated as three 1-D parallel channels.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
See Figure B-13.
c) Please provide in detail the nodalisation of the steam line.
See Figure B-13.
Figure B-13. NUPEC Peach Bottom 2 plant nodalisation

106

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Core flow rate and feedwater at initial were from the final specifications. Feedwater flow rate andtemperature during the transient were also from the final specifications.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Fission power was from the final specifications. Decay heat was taken in account.

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
Core bypass region was simulated by one vessel region. The core bypass flow was simulated witha leak path model from the T-H channel to the vessel (core bypass) region.
b) How are the jet pumps modelled?
The jet pumps were modelled with the JETP component of TRAC-BF1, following Table 3.1.2.2
and Figure 3.1.2.1 in the final specifications.
c) How is the steam separator modelled?
The steam separators were modelled with SEPD component using a mechanistic model of
TRAC-BF1. Code defaults were used.
d) How are the turbine stop valve closing and steam bypass system modelled?
The turbine stop valve and steam bypass valve were modelled with the VALVE component of
TRAC-BF1. The valve movements were given as timetables.
e) What were the difficulties encountered during the component design modelling?
Simulation of inertia of the steam separators.

4.

Generala) Deviations from the updated final specifications?
None.
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
Timesteps: 0.5-1.0 ms.
c) Specific features of the codes used?
None.

107

d) Number of solutions submitted per participant and how they differ?
One.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
Our result was within 3σ of the measured result.
NUPEC used the TRAC-BF1 code in the first exercise. Peach Bottom 2 plant nodalisation isshown above in Figure B-13. The thermal-hydraulic channels were treated as three 1-D parallel
channels. The reactor pressure vessel is modelled in 3-D cylindrical co-ordinates with two radial
rings and 15 axial levels.

108

RETRAN-3D (PSI, Switzerland)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
A 1-D nodalisation with 53 volumes and 63 junctions is used. The core region between the lower
plenum and the upper plenum is represented by a core inlet volume, a core exit volume, a core
bypass volume and a T-H channel representing the core, which consists of 24 axially stacked
volumes with pre-defined power.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
For the modelling of the transient, we used the code RETRAN-3D. Although the reference
solution for the benchmark was developed using RETRAN-02 [by Andy Olson (Exelon/USA)] in
order to fit the results as much as possible, we decided to develop a PSI RETRAN-3D model. The
aim was to increase our team’s learning curve. Our RETRAN-3D input model for the phase 1
calculation was developed from an early RETRAN-02 model for Turbine Trip 1 [Hornyik andNaser, 1979]. From this model, we used the nodalisation for the reactor vessel (including thesteam separator and downcomer region), the recirculation loop and the steam lines. As
preparation for the inclusion of the 3-D core in the later benchmark phases, the reactor coreregion was renodalised (including the accompanying heat structures) to include 24 axial control
volumes, in combination with a core exit volume and core inlet volume connected to the core
bypass volume. From the benchmark specifications, we included the time-dependence and axial
profile of the pre-defined power, together with the feedwater flow, the turbine inlet flow rate andthe turbine bypass area of Turbine Trip 2. In addition, the value of enthalpy in the lower plenum,
steam dome pressure and flow rates at t = 0 were used.
We investigated the sensitivity of our model with respect to non-equilibrium effects in the steam
separator region and to pressure losses in the turbine bypass line. The improvements with respect
to the non-equilibrium effects were included in the 3-D model, since they influence the magnitude
and timing of the power maximum. Improvements relating to pressure losses in the turbine bypass
line (effective only after t = 2 s) were not included.

109

Figure B-14. PSI nodalisation of the reactor vessel
To steamline

Steam dome dryer

Lower dryer

Feedwater

Steam sep. ext.

Steam separator

Upper downcomer

Standpipes

Lower downc.

Upper plenum

Recirculationlines
Jet pumps

CoreregionLower plenum

110Figure B-15. PSI nodalisation of the channels
To Standpipes
Upper plenumCore
exit vol.

Core,
1 TH channel,
24 volumes

CorebypassCore inlet
vol.
From
Jet pumps

Lower Plenum

Nodalisation of the core region: 1-D representation of the core, core inlet volume + core exit
volume + bypass volume + 1-D representation of the core consisting of 24 axially stacked
volumes levels, which are connected to “heating elements” with pre-defined power.
c) Please provide in detail the nodalisation of the steam line.
Figure B-16 gives the detailed nodalisation of the steam line.
2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
At time zero, the liquid enthalpy in the lower plenum and the system pressure in the steam dome
are defined. Also specified at time zero are the total flow rate through the core, the flow rate fromthe lower plenum to the core bypass and the flow rate from the core inlet volume to the core
bypass volume, which define all flow rates in the core region and give relevant loss coefficients at
the junctions. The loss coefficients at the junctions used at time zero are then also applied by the
RETRAN-3D code for all other times.

111b) Which neutronic/power initial and transient boundary conditions are used and how?
Twenty-four (24) axially stacked volumes are used in the core with pre-defined power. The axial
power shape in the core is time-independent and the time-averaged power of the core is definedas described in the final specifications.
Figure B-16. PSI nodalisation of the steam line
From steam
dome
54.875 ft

Safety Valves
STM L INE to SR/V
1056.15 ft3
14
L=83.177 ft
J18
12.698 ft2

8.3646 ft

15

SR/V to S TM ISOV
398.21 ft3
L=31.361 ft-10.083 ft

16
5.5416 ft

27
5.25 ft

17

STM BPS Chest
211.94 ft3
L=70.170 ft

5.6875 ft
BPS valves

31

J20 5.0000 ft
3.0204 ft2
5.2552 ft

29

STM BPS Lines
608.43 ft3
L=215.72 ft-32.224 ft

32J270.6753 ft2

28

4.25 ft

J41
0.6842 ft2

STM BPS Orifice16.719 ft3
L=3.0417 ft

J19
12.698 ft2

Vol. 16-A-1
945.51 ft3
L=74.464 ft
J34
12.698 ft2
Vol. 16-A-2
1294.4 ft3
L=101.94 ft
J35
12.698 ft2
Vol. 16-B-1
979.70 ft3
L=77.156 ft
J36
12.698 ft2
Vol. 16-B-2
1204.68 ft3
L=94.875 ft
J22
1.0 ft2
To turbine

J42
1.8571 ft2-32.224 ft

33

112

Condenser
1.0E+9 ft3
L=1.0 ft

3 S/RV’s

3. Component design and modellinga) How many core bypass channels are used in this exercise?
One core bypass channel.
b) How are the jet pumps modelled?
The jet pumps are modelled using a RETRAN-3D pump model.
c) How is the steam separator modelled?
In order to determine the distribution of the steam flow, it is important to follow the flow of steam
through the separators into the steam dome. In the reactor, two-phase flow from the core region
passes through the standpipes and then enters the steam separators. In each separator, thesteam-water mixture passes turning vanes, which impart a spin to establish a vortex and separates
the water from the steam. The denser liquid is thrown radially outward by centrifugal force
forming a continuous film on the inside wall of the inner pipe. The separator water exits fromunder the separator cap and flows out between the standpipes, draining into the downcomer.
Steam with a quality of at least 90% exits from the top of the separator and rises to the dryers.
The dryers force the wet steam to be directed horizontally through the dryer panels. The steam
flow makes a series of rapid changes in direction while traversing the panels. During these
direction changes, the heavier drops of entrained moisture are forced to the outer walls where
moisture collection hooks catch and drain the liquid to collection troughs, then through tubes into
the vessel downcomer so that the steam dryer assembly increases the quality of the steam to morethan 99.9%. This dry steam flows into the steam dome and the steam lines.
This behaviour is approximated in the RETRAN-3D model by control volumes (CV 1-5 inFigure B-17). Two-phase flow from the core region flows into the steam separator volume (CV 1).
In this volume, a liquid level is defined, below which a two-phase flow is established. Above theliquid level there is only steam flow. In this volume (CV 1), the RETRAN-3D bubble rise model
option is used. The steam flows upwards to the lower dryer volume (CV 2) and continues to thesteam dome volume (CV 3) and into the steam line.
Two-phase flow, which consists of water with some vapour carry-under, flows from the steam
separator volume to the steam separator external volume (CV 4). The junction between these twocontrol volumes is below the liquid level in CV 1. In the steam separator external (CV 4), a liquid
level develops with a steam phase above the liquid level and a two-phase region below the liquid
level. The RETRAN-3D bubble rise model is again used to describe this phenomenon. At t = 0
there is a very small steam flow from the steam separator external to the lower dryer volume(CV 2), while the main flow is water down into the upper downcomer (CV 5). The feedwater also
flows into the upper downcomer volume. From the upper downcomer volume, water flows to thelower downcomer and jet pumps, and from there back into the core.
Since the RETRAN-3D code option used in analysis is that of the four-equation model (i.e. fullthermal equilibrium), the volume of water in steam separator (CV 1) and steam separator external(CV 4) two-phase control volumes is in equilibrium with the steam. And, this strongly influences
the pressure increase per unit of steam generated in the core. As a result, a renodalisation of theseparator/downcomer region was performed in a sensitivity study, which gave a pressure increase
equal to the measured one at ~1 s. In order to achieve this, the volumes of the steam separator andthe steam separator external control volumes were reduced in the sensitivity test, while thevolume of the upper downcomer control volume was increased to preserve the total water volume.
113

Figure B-17. Nodalisation of the steam separator region
to SteamLine

Steam Dome Dryer, CV 3

Lower Dryer, CV 2

Feed
Water

Steam Sep. Ext.,
CV 4

Steam Separator,
CV 1

Upper Downcomer,
CV 5from Standpipes

to LowerDowncomer“Core”

to JetPumps

The main effect of renodalisation is a reduction in energy transfer from steam to liquid, which isdue to the volume being smaller where thermal equilibrium is established. Thus, there is lesscondensation of the vapour as the pressure increases. To obtain the measured pressure increase, it
was necessary to approximately halve the volume of the steam separator and the steam separator
external control volumes.
Another RETRAN-3D code option that can be used to obtain a reduced energy transfer between
liquid and vapour is the “two-region non-equilibrium model”. (Note that this model was used inconjunction with the RETRAN-3D bubble rise model to obtain the submitted results.) This model
divides the fluid in the control volume at the liquid level interface into two regions, where eachregion is in internal thermal equilibrium but the two regions are not necessarily in equilibriumwith each other. The liquid region below the interface and the vapour region above the interface
have the same pressure, but in general have different temperatures (i.e. in the vapour region we
may have superheated steam). The application of this model (in the steam separator volume andthe steam separator external volume, CVs 1 and 4 in Figure B-17) also produced an increased
pressure at ~1 s. A sensitivity study showed that the effect in the steam separator volume isnegligible and the increased pressure at 1 s comes mostly from the steam separator external
volume. The reason for this phenomenon is that after 0.3 s in the steam separator external volume,the pressure increase leads to an increase in the steam temperature in the vapour region. However,
due to the non-equilibrium model, the energy transfer between the vapour region and the liquid
region is reduced, maintaining the superheated steam in the vapour and dryer regions. In thesteam separator region, there is a transfer of vapour from the liquid region at the level interface
because of rising bubbles through the liquid region. Related to this phenomenon, the reduction ofenergy transfer across the interface due to the non-equilibrium model generates only a small effect.

114d) How are the turbine stop valve closing and steam bypass system modelled?
The turbine stop valve closing is described by a negative fill junction model, which is combinedwith a valve where area versus time is specified. This defines the flow rate to the turbine.
The steam bypass system is modelled using four RETRAN-3D control volumes. The steam
bypass chest volume is connected via the bypass valves to the steam bypass lines volume, whichis connected to the steam bypass orifice volume that is in turn connected to the condenser. In theturbine bypass valve between the steam bypass chest volume and the steam bypass lines volume,area versus time is defined (approximately) to get the specified flow through the turbine bypass.
Choked flow through the bypass valve is assumed.
e) What were the difficulties encountered during the component design modelling?

4.

Generala) Deviations from the updated final specifications?
Since we decided that there would be no point just copying the reference RETRAN-02 model, our
PSI RETRAN-3D model must have some differences in the final specifications [see also response
to question 1b)]. For instance, the nodalisation of steam line, reactor vessel and steam separator
region is based on the original model from Hornyik and Naser (1979). This means, for example,
that the steam separator volume is connected to the steam separator external volume. That is to
say that the two RETRAN-3D volumes with a liquid level are connected, which is not the case in
the Olson model.
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
The minimum timestep size between 0-1.2 s is 0.001 s and 0.006 s after 1.2 s. However, asensitivity study showed that the effect of choosing an overall minimum timestep of 0.006 s
would have only a negligible effect on the results.
c) Specific features of the codes used?
As described above, the four-equation model was used for the two-phase flow calculation, together
with an algebraic (Zolotar-Lellouche) slip correlation. The bubble rise model was used in the steam
separator and steam separator external volumes to describe the development of a liquid level. The
two-region non-equilibrium model was also used in the steam separator and steam separator
external volumes to account for the relatively slow exchange of energy across the liquid level.
d) Number of solutions submitted per participant and how they differ?
One solution from PSI.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

115

TRAC-BF1 (PSU, USA)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
Three-dimensional (3-D) nodalisation for VESSEL, 1-D nodalisation for other components.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
See the PSU TRAC-BF1 nodalisation figure that follows (Figure B-18).
c) Please provide in detail the nodalisation of the steam line.
Straight forward nodalisation is used for steam line as shown as in Figure B-18 below.
Figure B-18. PSU TRAC-BF1 nodalisation

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Same as the boundary conditions in specifications.
116b) Which neutronic/power initial and transient boundary conditions are used and how?
Power lookup table from specifications is used.
3.

Component design and modellinga) How many core bypass channels are used in this exercise?
No bypass channel is used. Core bypass flow is simulated by leak paths from first cell of channels
to related elevation of the vessel.
b) How are the jet pumps modelled?
Standard jet pump option of TRAC-B is used. See Figure B-18.
c) How is the steam separator modelled?
Simple separator is used.
d) How are the turbine stop valve closing and steam bypass system modelled?
Same as specifications.
e) What were the difficulties encountered during the component design modelling?
No difficulties were encountered.

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
The convergence criterion is 0.0001 and courant number is 0.999. The timestep size is 6 ms.
Some loss coefficients are assumed.
c) Specific features of the codes used?
See the TRAC-BF1 code description in Appendix A.
d) Number of solutions submitted per participant and how they differ?
One solution is provided.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
The results agree well with the provided reference results.
117

TRAC-M (PSU/NRC, USA)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
Three-dimensional (3-D) nodalisation for VESSEL, 1-D nodalisation for other components.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
See Figure B-19.
c) Please provide in detail the nodalisation of the steam line.
Straight forward nodalisation is used for steam line as shown as in Figure B-19.

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Same as the boundary conditions in specifications.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Power lookup table from specifications is used.

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
No bypass channel is used.
b) How are the jet pumps modelled?
Standard jet pump option of TRAC-M is used. See Figures B-19 and B-20.
c) How is the steam separator modelled?
Three-stage mechanistic separators are used.
d) How were the turbine stop valve closing and steam bypass system modelled?
They are the same as specifications.

118e) What were the difficulties encountered during the component design modelling?
No difficulties were encountered.
Figure B-19. PSU/NRC TRAC-M nodalisation

119Figure B-20. PSU/NRC TRAC-M jet pump nodalisation

Figure B-21. PSU/NRC TRAC-M separator nodalisation

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
The convergence criterion is 0.0001 and courant number is 0.999. The timestep size is 6 ms.
Some loss coefficients are assumed.
c) Specific features of the codes used?
See the TRAC-M code description in Appendix A.
d) Number of solutions submitted per participant and how they differ?
One solution is provided.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
Generally, the results do not differ from the reference results (especially for first 1.5 s).
120

TRAC-BF1 (TEPSYS, Japan)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
VESSEL 3-D, others 1-D, core T-H based on specifications.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
See the TEPSYS TRAC-BF1 plant model that follows (Figure B-22).
c) Please provide in detail the nodalisation of the steam line.
This nodalisation is common for most transient cases.
Figure B-22. TEPSYS TRAC-BF1 plant model

121

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Based on specifications.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Based on specifications.

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
Two.
b) How are the jet pumps modelled?
Single JETP model based on specifications and other aspects based on ordinary practice.
c) How is the steam separator modelled?
Single SEPD model based on specifications, other aspects on ordinary practice (simple separator).
d) How are the turbine stop valve closing and steam bypass system modelled?
Strictly based on specifications. Valve area was derived based on semi-empirical correlation.
e) What were the difficulties encountered during the component design modelling?

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
Delta-t was controlled automatically. The courant number was specified as unity.
c) Specific features of the codes used?
TRAC-BF1.
d) Number of solutions submitted per participant and how they differ?
It appears that we submitted only one solution as the reference condition. Other results, if any,must be regarded as done within the scope of the parametric study.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
122

RELAP5 (UPISA, Italy)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
One-dimensional (1-D) for the plant loop and 33 heated core channels, and a total of 1 230
volumes and 1 267 junctions.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
See Figure B-23 for a schema of the adopted nodalisation.
c) Please provide in detail the nodalisation of the steam line.
Table B-1. UPISA steam line and bypass system geometrical data
Item

Length (m)Flow area (m2)

Steam line 1
Steam line 2
Steam line 3
Steam line 4
Steam line 5
Steam line 6
Steam line 7
Steam bypass 8
Steam bypass valve
Steam bypass 9

18.0
114.6831.5
114.3646.4
110.22
29.1
6.9
–
51.7

1.133
1.133
1.133
1.133
1.133
1.133
1.133
10.2767
10.0613
10.2650123

Equivalentdiameter (m)
0.6
0.6
0.6
0.6
0.6
0.6
0.6
110.403–110.194

Figure B-23. UPISA RELAP5 model

Figure B-24. UPISA steam line simplified schema

124

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
No boundary conditions are specified in the core zone.
b) Which neutronic/power initial and transient boundary conditions are used and how?
Table B-2. UPISA decay constant and fractions of delayed neutrons
Group

Decay constant (s–1)

1
2
3
4
5
6

0.012813
0.031536
0.124703
0.328273
1.405280
3.844728

Relative fraction of delayed
neutrons in %0.01670.11340.10220.21520.08370.0214

Total fraction of delayed neutrons equalled 0.5526%.
Decay heat using ANS-79 standard fission product data.
Transient reactivity table as specified in the BWR TT benchmark specification CD-ROM.
3.

Component design and modellinga) How many core bypass channels are used in this exercise?
One.
b) How are the jet pumps modelled?
The same as in the BWR TT benchmark specifications.
c) How is the steam separator modelled?
The default option of the RELAP5 separator.
d) How are the turbine stop valve closing and steam bypass system modelled?
Linear TSV closure.
The steam bypass system model is shown in Figure B-24 where the condenser is a big volume
with thermal-hydraulic conditions of pressure equal to 17.5 MPa and equilibrium steam quality
equal to 0.99e) What were the difficulties encountered during the component design modelling?
To get the right individual channels steady state mass flow rate for the orifices at core inlet.
125

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
Minimum DT = 1.0 e – 6 s
Maximum DT = 1.0 e – 3 sc) Specific features of the codes used?
The RELAP5/MOD3 code is based on a six-equation non-homogeneous and non-equilibrium
model for the two-phase system. The solutions are obtained through fast resolution of a partially
implicit numerical scheme.
d) Number of solutions submitted per participant and how they differ?
Two solutions, the main differences between them are outlined in the table below.

First solution
Second solution

Number ofheated channels
133Core pressure
drop MPa
0.06700.1025

Inlet enthalpy
1 194.0
1 208.5e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

126

TRAC-BF1-VALKIN (UPV, Spain)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
The nodalisation used is 3-D and we have used 33 T-H channels.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
Figure B-25. UPV thermal-hydraulic nodalisation
BREAK

VESSEL
SRV

TSV
SEPD

BREAK
TEE

TEE

33 CHANNELS
26

BYPASS VALVE

BREAK
CHAN

JETPUMP

PUMP
4
3
2
1
FILL

127

Figure B-26. UPV thermal-hydraulic channel map

0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0
0 0 18 17 17 17 17 17 17 17 17 17 17 17 17 18 0 0
0 0 0 33 14 15 14 15 14 15 15 15 15 14 15 14 15 14 33 0 0 0
0 33 18 16 15 14 15 14 15 14 14 14 14 15 14 15 14 15 16 18 33 0
0 0 33 14 15 13 13 13 13 13 11 13 13 11 13 13 13 13 13 15 14 33 0 0
0 0 0 18 14 15 13 12 13 12 11 12 11 10 10 11 12 11 12 13 12 13 15 14 18 0 0 0
0 33 33 29 26 11 13 11 13 11 13 13 11 11 11 11 13 13 11 13 11 13 11 26 29 33 33 0
0 0 18 29 30 27 26 11 12 11 12 7 6 7 8 8 7 6 7 12 11 12 11 26 27 30 29 18 0 0
0 33 29 30 13 13 27 25 7 7 7 7 7 8 8 8 8 7 7 7 7 7 25 27 13 13 30 29 33 0
17 29 30 13 28 27 28 22 31 7 31 7 6 7 6 6 7 6 7 31 7 31 22 28 27 28 13 30 29 17
17 30 29 13 27 13 27 24 22 23 7 8 3 3 3 3 3 3 8 7 23 22 24 27 13 27 13 29 30 17
17 29 30 27 28 27 28 22 31 22 4 3 4 3 3 3 3 4 3 4 22 31 22 28 27 28 27 30 29 17
17 30 29 27 13 13 22 22 22 24 21 5 3 3 3 3 3 3 5 21 24 22 22 22 13 13 27 29 30 17
17 29 30 27 28 27 32 22 32 21 4 21 4 19 2 2 19 4 21 4 21 32 22 32 27 28 27 30 29 17
17 30 29 27 27 27 24 22 22 20 20 20 19 19 1 1 19 19 20 20 20 22 22 24 27 27 27 29 30 17
17 30 29 27 10 9 22 22 32 20 20 20 2 1 1 1 1 2 20 20 20 32 22 22 9 10 27 29 30 17
17 30 29 27 10 9 22 22 32 20 20 20 2 1 1 1 1 2 20 20 20 32 22 22 9 10 27 29 30 17
17 30 29 27 27 27 24 22 22 20 20 20 19 19 1 1 19 19 20 20 20 22 22 24 27 27 27 29 30 17
17 29 30 27 28 27 32 22 32 21 4 21 4 19 2 2 19 4 21 4 21 32 22 32 27 28 27 30 29 17
17 30 29 27 13 13 22 22 22 24 21 5 3 3 3 3 3 3 5 21 24 22 22 22 13 13 27 29 30 17
17 29 30 27 28 27 28 22 31 22 4 3 4 3 3 3 3 4 3 4 22 31 22 28 27 28 27 30 29 17
17 30 29 13 27 13 27 24 22 23 7 8 3 3 3 3 3 3 8 7 23 22 24 27 13 27 13 29 30 17
17 29 30 13 28 27 28 22 31 7 31 7 6 7 6 6 7 6 7 31 7 31 22 28 27 28 13 30 29 17
0 33 29 30 13 13 27 25 7 7 7 7 7 8 8 8 8 7 7 7 7 7 25 27 13 13 30 29 33 0
0 0 18 29 30 27 26 11 12 11 12 7 6 7 8 8 7 6 7 12 11 12 11 26 27 30 29 18 0 0
0 33 33 29 26 11 13 11 13 11 13 13 11 11 11 11 13 13 11 13 11 13 11 26 29 33 33 0
0 0 0 18 14 15 13 12 13 12 11 12 11 10 10 11 12 11 12 13 12 13 15 14 18 0 0 0
0 0 33 14 15 13 13 13 13 13 11 13 13 11 13 13 13 13 13 15 14 33 0 0
0 33 18 16 15 14 15 14 15 14 14 14 14 15 14 15 14 15 16 18 33 0
0 0 0 33 14 15 14 15 14 15 15 15 15 14 15 14 15 14 33 0 0 0
0 0 18 17 17 17 17 17 17 17 17 17 17 17 17 18 0 0
0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0c) Please provide in detail the nodalisation of the steam line.
Figure B-27. UPV steam line nodalisation
BREAK

SRV
VESSELØ1.13
0.2

LEVEL 7

TSV

Ø0.16
24.09

22.53

14.53

14.53

BREAK
15.84

14.86

26.59

TEE

TEE

Ø1.13

22.10
Ø0.32

BYPASS
VALVE
BREAK

128

0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
We used the boundary conditions from specifications.
b) Which neutronic/power initial and transient boundary conditions are used and how?
The initial power was 2 030 MW, and the boundary conditions were from specifications.

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
We did not use any bypass channels.
b) How are the jet pumps modelled?
The jet pumps are modelled with the jet pump component of TRAC-BF1. The primary tube has
four cells and secondary one has two cells.
c) How is the steam separator modelled?
The steam separator was modelled with the TRAC-BF1 component SEPD.
d) How are the turbine stop valve closing and steam bypass system modelled?
The turbine valve closes at 0.096 s; it is specified in table provided by specifications on valve
components.
The steam bypass system is modelled using the data from specifications.
e) What were the difficulties encountered during the component design modelling?

4.

Generala) Deviations from the updated final specifications?
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
Timestep size of 0.1 s.
c) Specific features of the codes used?
d) Number of solutions submitted per participant and how they differ?
We submitted one solution.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.
129

POLCA-T (Westinghouse, Sweden)

1.

Thermal-hydraulic model and nodalisation of the vessel and balance of the planta) What kind of nodalisation is used (1-D, 3-D and number of T-H channels or cells)?
Three-dimensional (3-D) nodalisation with 764 T-H channels, 24 fuel nodes and two nodes forthe bottom and top reflector in each channel.
b) How are the volumes, junctions and channels/T-H cells chosen? (If possible, please provide
schematic.)
One T-H channel per fuel assembly, 122 radial reflector channels.
c) Please provide in detail the nodalisation of the steam line.
Main steam lines are modelled by steam lines, safety and relieve valves, turbine stop valves andsteam head with a total of 65 nodes and five valves.

2.

Boundary conditionsa) Which core thermal-hydraulic initial and transient boundary conditions are used and how?
Power versus timetable, turbine pressure controller set point versus time, bypass valve position
versus time and feedwater mass flow.
b) Which neutronic/power initial and transient boundary conditions are used and how?
See response to question 2a).

3.

Component design and modellinga) How many core bypass channels are used in this exercise?
One.
b) How are the jet pumps modelled?
The POLCA-T jet pumps model is used, which is validated against full and 1/6 scale test facility
data and has nodalisation similar to that used in TRAC-M calculations.
c) How is the steam separator modelled?
The POLCA-T steam separator model is used, which is validated against our test data.

130d) How are the turbine stop valve closing and steam bypass system modelled?
The turbine stop valve closing is modelled by a signal-activated turbine pressure controller. The
controller’s set point versus timetable is given as a boundary condition.
Bypass valve position versus time is given as a boundary condition. The steam bypass systemmodel covers: bypass chest, bypass valves, lines and orifice, and steam condenser with a total of
51 nodes and one valve.
d) What were the difficulties encountered during the component design modelling?
There were no difficulties encountered during the component design modelling. However, to set
up the plant model there was the need for a sensitivity study on some of the parameters [such as
carry-under and carry-over, turbine stop valve closing (linear or non-linear model), and jet pumpmodel to work along the M-N curve].
4.

Generala) Deviations from the updated final specifications?
We did not see any.
b) User assumptions? Please provide the timestep(s) size or maximum and minimum values used
for the simulation of Exercise 1.
Constant timestep as given in the specifications (0.006 s).
c) Specific features of the codes used?
Core description given in the input data of our neutronics code POLCA-7, POLCA-T steam
separator model, POLCA-T jet pumps model, control rods speed and position controller, scram
controller, jet pump drive flow controller, feedwater controller, RPV water level controller andturbine pressure controller.
d) Number of solutions submitted per participant and how they differ?
One solution.
e) If your results are significantly different (> 3σ) than the measured or Exelon results, arethere any explanations for these differences? Please explain each significant difference.

131OECD PUBLICATIONS, 2 rue André-Pascal, 75775 PARIS CEDEX 16
Printed in France.