Thermohydraulic relationships for advanced water cooled reactors

352 Pages·2001·12.979 MB·English

by  International Atomic Energy Agency

IAEA-TECDOC-1203

Thermohydraulic relationships for
advanced water cooled reactors

April 2001
The originating Section of this publication in the IAEA was:
Nuclear Power Technology Development Section
International Atomic Energy Agency
Wagramer Strasse 5
P.O. Box 100
A-1400 Vienna, Austria

THERMOHYDRAULIC RELATIONSHIPS FOR ADVANCED
WATER COOLED REACTORS
IAEA, VIENNA, 2001
IAEA-TECDOC-1203
ISSN 1011–4289
© IAEA, 2001
Printed by the IAEA in Austria
April 2001
FOREWORD

This report was prepared in the context of the IAEA’s Co-ordinated Research Project
(CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was
started in 1995 with the overall goal of promoting information exchange and co-operation in
establishing a consistent set of thermohydraulic relationships which are appropriate for use in
analyzing the performance and safety of advanced water cooled reactors. For advanced water
cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post
CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat
transport by natural circulation, condensation of steam in the presence of non-condensables,
thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow
instabilities.

Thirteen institutes co-operated in this CRP during the period from 1995 to 1999.

The IAEA acknowledges the strong efforts of the following persons in preparing this
report: N. Aksan (Paul Scherrer Institute, Switzerland), F. D'Auria (University of Pisa, Italy),
D.C. Groeneveld (AECL Research, Canada), P.L. Kirillov (Institute of Physics and Power
Engineering, Russian Federation) and D. Saha (Bhabha Atomic Research Centre, India). The
IAEA officers responsible for this publication were A. Badulescu and J. Cleveland of the
Division of Nuclear Power.
EDITORIAL NOTE
The use of particular designations of countries or territories does not imply any judgement by the
publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and
institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does
not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or
recommendation on the part of the IAEA.
CONTENTS

CHAPTER 1. INTRODUCTION..............................................................................................1
1.1. Overview of advanced water cooled reactors....................................................................1
1.2. Background for the co-ordinated research project............................................................4
1.3. Objectives..........................................................................................................................5
1.4. Participants.........................................................................................................................5
1.5. Summary of activities within the co-ordinated research project.......................................6
1.6. Structure of this report.......................................................................................................6
References to Chapter 1..............................................................................................................6

CHAPTER 2. THERMOHYDRAULIC PHENOMENA OF INTEREST TO
ADVANCED WATER COOLED REACTORS................................................9
References to Chapter 2............................................................................................................13
CHAPTER 3. A GENERAL CHF PREDICTION METHOD FOR
ADVANCED WATER COOLED REACTORS..............................................15
Nomenclature............................................................................................................................15
3.1. Introduction.....................................................................................................................16
3.2. CHF mechanisms.............................................................................................................18
3.2.1. General.................................................................................................................18
3.2.2. DNB (departure from nucleate boiling)...............................................................18
3.2.3. Helmholtz instability............................................................................................19
3.2.4. Annular film dryout.............................................................................................19
3.2.5. Unstable or periodic dryout.................................................................................20
3.2.6. Slow dryout..........................................................................................................20
3.3. CHF database...................................................................................................................20
3.3.1. General.................................................................................................................20
3.3.2. Tube database.......................................................................................................22
3.3.3. Bundle database...................................................................................................24
3.4. CHF prediction methodology..........................................................................................24
3.4.1. General.................................................................................................................24
3.4.2. Analytical models................................................................................................26
3.4.3. Empirical CHF prediction methods.....................................................................26
3.4.4. Application to bundle geometries........................................................................28
3.5. Recommended CHF prediction method for advanced water-cooled reactors.................30
3.5.1. Tubes....................................................................................................................30
3.5.2. Rod bundles.........................................................................................................31
3.5.3. Correction factors.................................................................................................31
3.6. Assessment of accuracy of the recommended prediction methods.................................37
3.6.1. CHF look up table assessment.............................................................................37
3.6.2. Accuracy of bundle CHF prediction methods.....................................................37
3.6.3. Impact of accuracy of CHF model on cladding temperature prediction..............39
3.7. CHF concerning accident conditions...............................................................................39
3.7.1. General.................................................................................................................39
3.7.2. Effect of the axial/radial node size.......................................................................39
3.7.3. Transient effects on CHF.....................................................................................42
3.8. Recommendations and final remarks..............................................................................42
References to Chapter 3............................................................................................................43
CHAPTER 4. GENERAL FILM BOILING HEAT TRANSFER
PREDICTION METHODS FOR ADVANCED
WATER COOLED REACTORS.....................................................................49
Nomenclature............................................................................................................................49
4.1. Introduction.....................................................................................................................50
4.2. Description of post-CHF phenomena..............................................................................52
4.2.1. General.................................................................................................................52
4.2.2. Transition boiling.................................................................................................53
4.2.3. Minimum film boiling temperature.....................................................................55
4.2.4. Flow film boiling.................................................................................................56
4.3. Film boiling data base.....................................................................................................60
4.3.1. General.................................................................................................................60
4.3.2. Tube and annuli....................................................................................................61
4.3.3. Bundle..................................................................................................................61
4.4. Overview of film boiling prediction methods.................................................................61
4.4.1. General.................................................................................................................61
4.4.2. Pool film boiling equations..................................................................................66
4.4.3. Film boiling models.............................................................................................70
4.4.4. Flow film boiling correlations.............................................................................75
4.4.5. Look-up tables for film boiling heat transfer in tubes.........................................86
4.5. Recommended/most recent film boiling prediction methods..........................................88
4.5.1. Pool film boiling..................................................................................................88
4.5.2. Flow film boiling.................................................................................................89
4.5.3. Radiation heat transfer in film boiling.................................................................90
4.5.4. Correlations for single phase heat transfer to superheated steam........................91
4.5.5. Application to rod bundles...................................................................................91
4.6. Application to film boiling prediction methods codes....................................................93
4.6.1. General.................................................................................................................93
4.7. Conclusions and final remarks........................................................................................94
References to Chapter 4............................................................................................................95
CHAPTER 5. PRESSURE DROP RELATIONSHIPS.........................................................109
Nomenclature..........................................................................................................................109
5.1. Introduction...................................................................................................................110
5.2. Survey of situations where pressure drop relationships are important..........................111
5.2.1. Distinction between core and system approach.................................................113
5.2.2. Geometric conditions of interest........................................................................113
5.3. Correlations for design and analysis .............................................................................115
5.3.1. Components of pressure drop............................................................................115
5.3.2. Configurations....................................................................................................117
5.3.3. Friction pressure drop correlations....................................................................118
5.3.4. Local pressure drop............................................................................................128
5.3.5. Importance of void fraction correlations............................................................132
5.3.6. Review of previous assessments........................................................................134
5.3.7. Proposed assessment procedure for diabatic vertical flow................................142
5.3.8. Results of assessment.........................................................................................142
5.4. Comparisons of correlations as they stand in codes......................................................146
5.4.1. Physical models in system codes.......................................................................147
5.5. Final remarks.................................................................................................................149
References to Chapter 5..........................................................................................................151
CHAPTER 6. REMARKS AND FUTURE NEEDS.............................................................163
References to Chapter 6..........................................................................................................165
APPENDICES I–XIX

APPENDIX I: ACTIVITIES CONTRIBUTED TO THE CRP BY THE
RESEARCH GROUPS AT THE PARTICIPATING INSTITUTES....169

APPENDIX II: THE 1995 LOOK-UP TABLE FOR
CRITICAL HEAT FLUX IN TUBES ..................................................175

APPENDIX III: CHF PREDICTION FOR WWER-TYPE
BUNDLE GEOMETRIES.....................................................................183

APPENDIX IV: AECL LOOK-UP TABLE FOR
FULLY DEVELOPED FILM-BOILING
HEAT-TRANSFER COEFFICIENTS (kW m-2 K-1)............................190

APPENDIX V: IPPE TABLE OF HEAT TRANSFER COEFFICIENTS FOR
FILM BOILING AND SUPERHEATED STEAM FOR TUBES........241

APPENDIX VI: CIAE METHOD FOR DETERMINING
FILMBOILING HEAT TRANSFER....................................................271

APPENDIX VII: TWO-PHASE VISCOSITY MODELS FOR USE
IN THE HOMOGENEOUS MODEL FOR
TWO-PHASE PRESSURE DROP........................................................273

APPENDIX VIII: TWO-PHASE PRESSURE DROP CORRELATIONS
BASED ON THE MULTIPLIER CONCEPT.......................................274

APPENDIX IX: DIRECT EMPIRICAL TWO-PHASE
PRESSURE DROP CORRELATIONS.................................................281

APPENDIX X: FLOW PATTERN SPECIFIC PRESSURE DROP
CORRELATIONS FOR HORIZONTAL FLOW....................................283

APPENDIX XI: FLOW PATTERN SPECIFIC PRESSURE DROP FOR
VERTICAL UPWARD FLOW.............................................................288

APPENDIX XII: INTERFACIAL FRICTION MODELS
GIVEN BY SOLBRIG (1986)...............................................................291
APPENDIX XIII: SLIP RATIO MODELS FOR CALCULATION OF
VOID FRACTION.................................................................................294
APPENDIX XIV: K(cid:1) MODELS FOR THE CALCULATION OF
VOID FRACTION.................................................................................295

APPENDIX XV: DRIFT FLUX MODELS FOR THE CALCULATION OF
VOID FRACTION.................................................................................296

APPENDIX XVI: MISCELLANEOUS EMPIRICAL CORRELATIONS FOR
VOID FRACTION.................................................................................304

APPENDIX XVII: COMPILATION OF DATA..................................................................305

APPENDIX XVIII: DETAILED RESULTS OF ASSESSMENT OF
VOID FRACTION CORRELATIONS.................................................310

APPENDIX XIX: DETAILED RESULTS OF ASSESSMENT OF
FLOW PATTERN DATA.....................................................................319

ANNEX A: INTERNATIONAL NUCLEAR SAFETY CENTER DATABASE................331
ANNEX B: PREPARED METHODOLOGY TO SELECT RANGES OF
THERMOHYDRAULIC PARAMETERS........................................................335
CONTRIBUTORS TO DRAFTING AND REVIEW............................................................343
Chapter 1
INTRODUCTION

1.1. OVERVIEW OF ADVANCED WATER COOLED REACTORS

In the second half of the 20th century nuclear power has evolved from the research and
development environment to an industry that supplies 17% of the world’s electricity. In these
50 years of nuclear development a great deal has been achieved and many lessons have been
learned. At the end of 1998, according to data reported in the Power Reactor Information
System, PRIS, of the IAEA, there were 434 nuclear power plants in operation and 34 under
construction. Over eight thousand five hundred reactor-years of operating experience had
been accumulated.

Due to further industrialization, economic development and projected increases in the
world’s population, global energy consumption will surely continue to increase into the 21st
century. Based on IAEA’s review of nuclear power programmes [IAEA (1998)] and plans of
Member States, several countries, especially in the Far East, are planning to expand their
nuclear power capacity considerably in the next 15–20 years.

The contribution of nuclear energy to near and medium term energy needs depends on
several key issues. The degree of global commitment to sustainable energy strategies and
recognition of the role of nuclear energy in sustainable strategies will impact its future use.
Technological maturity, economic competitiveness and financing arrangements for new plants
are key factors in decision making. Public perception of energy options and related
environmental issues as well as public information and education will also play a key role in
the introduction of evolutionary designs. Continued vigilance in nuclear power plant
operation, and enhancement of safety culture and international co-operation are highly
important in preserving the potential of nuclear power to contribute to future energy
strategies.

To assure that nuclear power remains a viable option in meeting energy demands in the
near and medium terms, new reactor designs, aimed at achieving certain improvements over
existing designs are being developed in a number of countries. Common goals for these new
designs are high availability, user-friendly features, competitive economics and compliance
with internationally recognized safety objectives.

The early development of nuclear power was to a large extent conducted on a national
basis. However, for advanced reactors, international co-operation is playing an important role,
and the IAEA promotes international co-operation in advanced reactor development and
application. Various organizations are involved, including governments, industries, utilities,
universities, national laboratories, and research institutes.

Worldwide there is considerable experience in nuclear power technology and especially
in light water reactor (LWR) and heavy water (HWR) technology. Of the operating plants,
346 are LWRs totaling 306 GW(e) and 31 are HWRs totaling 15 GW(e). The experience and
lessons learned from these plants are being incorporated into new water cooled reactor
designs. Utility requirements documents have been formulated to guide these activities by
incorporating this experience with the aim of reducing costs and licensing uncertainties by
establishing a technical foundation for the new designs.

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The full spectrum of advanced water cooled reactor designs or concepts covers different
types of designs — evolutionary ones, as well as innovative designs that require substantial
development efforts. A natural dividing line between these two categories arises from the
necessity of having to build and operate a prototype or demonstration plant to bring a concept
with much innovation to commercial maturity, since such a plant represents the major part of
the resources needed. Designs in both categories need engineering, and may also need
research and development (R&D) and confirmatory testing prior to freezing the design of
either the first plant of a given line in the evolutionary category, or of the prototype and/or
demonstration plant for the second category. The amount of such R&D and confirmatory
testing depends on the degree of both the innovation to be introduced and the related work
already done, or the experience that can be built upon. This is particularly true for designs in
the second category where it is entirely possible that all a concept needs is a demonstration
plant, if development and confirmatory testing is essentially completed. At the other extreme,
R&D, feasibility tests, confirmatory testing, and a prototype and/or demonstration plant are
needed in addition to engineering. Different tasks have to be accomplished and their
corresponding costs in qualitative terms are a function of the degree of departure from
existing designs. In particular, a step increase in cost arises from the need to build a reactor as
part of the development programme (see Figure 1.1).

Advanced design Evolutionary design
Different types of new nuclear plants are being developed An evolutionary design is an advanced design that achieves
today that are generally called advanced reactors. In improvements over existing designs through small to
general, an advanced plant design is a design of current moderate modifications, with a strong emphasis on
interest for which improvement over its predecessors and/or maintaining proven design features to minimise
existing designs is expected. Advanced designs consist of technological risks. The development of an evolutionary
evolutionary designs and designs requiring substantial design requires at most engineering and confirmatory
development efforts. The latter can range from moderate testing.
modifications of existing designs to entirely new design
concepts. They differ from evolutionary designs in that a Innovative design
prototype or a demonstration plant is required, or that An innovative design is an advanced design which
insufficient work has been done to establish whether such a incorporates radical conceptual changes in design
plant is required. approaches or system configuration in comparison with
existing practice. Substantial R&D, feasibility tests, and a
prototype or demonstration plant are probably required.

Figure 1.1. Efforts and development costs for advanced designs versus departure from
existing designs (Terms are excerpted from IAEA-TECDOC-936).
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