Universita’ di Pisa
Facoltà di Ingegneria
Corso di Laurea Specialistica in Ingegneria Energetica
Analysis of Flow Patterns in the I.R.I.S.
Reactor Downcomer and Lower Plenum
during Regular Operation
Candidato
Relatori
Stefano Gallo
prof. ing. Francesco Oriolo
prof. ing. Walter Ambrosini
dott. ing. Nicola Forgione
Milorad B. Dzodzo
Westinghouse Electric Co. LLC
Mario D. Carelli
Westinghouse Electric Co. LLC
Sessione di Laurea 8 Ottobre 2007 Anno Accademico 2006/2007
ai miei familiari
Aknowledgement
I would like to express my gratitude to all those who gave me the possibility to complete this thesis.
Particurarly, I want to thank my supervisors at the University of Pisa and my supervisors at the Westinghouse. They helped me and they gave me the opportunity to live a beautiful experience.
I would like to give my special thanks to my brother, my friends, my ”col-leagues” and my parents; their support enabled me to complete my studies.
Abstract
The challenge for Nuclear Industry in the 21st century is to achieve higher standards with regard to Sustainability, Economics, Safety and Reliability. This is the purpose of the Generation IV initiative. The International Reactor Innovative and Secure (IRIS), as a modular Light Water Reactor (LWR) based on existing proven technology, and with significantly improved safety, satisfies these requirements. The University of Pisa (Dipar-timento di Ingegneria Meccanica, Nucleare e della Produzione), as a member of the IRIS consortium, is involved in the Thermo-Hydraulic study and design of the IRIS core cooling system. The main objective of this activity is to contribute to the IRIS reactor design by optimizing the features of an experimental facility intended to provide data on reactor phenomena, and to validate the Computational Fluid Dynamic (CFD) code.
In the present work, a qualitative analysis, which has led to a Phenomena Identification and Ranking Table (PIRT), was performed on the IRIS reactor downcomer and lower plenum. By means of CFD, this analysis can be validated and quantified and some design issues can be addressed. Thus, the entrance of the coolant flow into the downcomer, one of the phenomena identified in the previous analysis, has been studied. The commercial code has been verified on an experimental reference case. Then the downcomer inlet has been modeled and four simulations have been completed. Two simulations were done on the full scale model, whereas other two simulations were done on a down scaled model in order to address the scaling analysis of the experimental facility.
Contents
1 Introduction 1
2 The Nuclear Industry in the 21st Century 3
2.1 Generation IV Technology Goals . . . 4
2.1.1 Sustainability . . . 4
2.1.2 Safety and Reliabity . . . 5
2.1.3 Economics . . . 5
2.2 The I.R.I.S. Project . . . 6
2.2.1 The I.R.I.S. Consortium . . . 6
2.2.2 IRIS Main Feature . . . 8
2.2.3 System Configuration . . . 8
2.2.4 Safety-by-Design Approach . . . 15
2.2.5 Optimized Maintenance . . . 19
2.2.6 Economics . . . 19
2.2.7 IRIS Status Assessment . . . 21
3 Scaling of Thermohydraulic Systems 23 3.1 The Hierarchical Two-Tiered Scaling Analysis . . . 24
3.1.1 System Decomposition and Hierarchy . . . 25
3.1.2 Top-down Approach . . . 31
3.1.3 Bottom-up Scaling Approach . . . 33
3.2 The Fractional Scaling Analysis . . . 35
3.2.1 Integral Approach to System Analysis . . . 35
3.2.2 Fractional Rate of Change . . . 36
3.2.3 Spatial and Time Scales . . . 36
3.2.4 Fractional Scaling Paradigm . . . 41
3.2.5 Application of Fractional Scaling Analysis to Nuclear Power Plant . 44 4 Thermohydraulic Analysis of the IRIS Downcomer and Lower Plenum. 53 4.1 The IRIS Downcomer and Lower Plenum . . . 53
4.2 Phenomena Identification and Ranking Table (PIRT) . . . 55
4.3 IRIS Downcomer and Lower Plenum: Thermo-hydraulic PIRT . . . 56
4.4 Design issues . . . 78
4.4.1 Steam Generator Outlet: Shape and Position . . . 78
4.4.2 Flow between the Diagonal and Bottom Shields . . . 79
4.4.3 Mixing Volume Layout . . . 86
5 Jet Mixing Phenomena and CFD Verification 89 5.1 Reference Case . . . 89
5.2 Previous CFD Validation . . . 92
5.3 Adopted Model and Calculation Cases . . . 93
5.4 Comparison between Experimental Data and Calculations . . . 98
5.4.1 Mesh Sensitivity Analysis . . . 99
5.4.2 Effect of Different Turbulence Models . . . 113
6 Downcomer Inlet Simulation 119 6.1 Adopted Model and Calculation Cases . . . 119
6.1.1 Geometry . . . 119
6.1.2 Computational Domain . . . 119
6.1.3 Calculation Cases . . . 124
6.1.4 Mesh . . . 126
6.1.5 Fluid Properties and Flow conditions . . . 126
6.2 Results . . . 128
6.2.1 Full Scale Simulations . . . 128
6.2.2 Down Scaled Simulations . . . 139
List of Figures
2.1 Compact confinement design in the IRIS project . . . 9
2.2 Axonometric view of the Reactor Pressure Vessel . . . 10
2.3 Cross section view of the Reactor Pressure Vessel . . . 11
2.4 Layout of the helical coil steam generators . . . 13
3.1 Flow diagram for the hierarchical two tiered analysis . . . 24
3.2 System decomposition and hierarchy . . . 26
3.3 Hierarchy . . . 27
3.4 Total transfer area concentration . . . 28
4.1 Cross section view of the downcomer and lower plenum . . . 54
4.2 Axonometric view of the downcomer and lower plenum . . . 54
4.3 Phenomena occurring within DC and LP. Bold font style indicates sub-regions 56 4.4 Flow between diagonal and bottom shield, wake generation. . . 66
4.5 Flow between diagonal and bottom shield: global and local effects. . . 67
4.6 Wake features. . . 67
4.7 Water jet features. . . 68
4.8 Steam generator possible shapes . . . 78
4.9 Steam generator possible positions . . . 79
4.10 Flow re-arrangement . . . 80
4.11 Cross section view of the diagonal shield . . . 80
4.12 Cross section view of the bottom shield . . . 81
4.13 Wake due to the circumferential velocity component . . . 81
4.14 Possible shapes of the supports . . . 82
4.15 Diagonal shield. Surfaces to be rounded . . . 83
4.16 Wake due to radial velocity component . . . 83
4.17 Improving the flow downstream the surface C . . . . 84
4.18 Shape improvement by rounding . . . 84
4.19 Cross flow area restriction between diagonal and bottom shield . . . 85 4.20 Proposal to increase the cross flow area between diagonal and bottom shield 85
4.21 Shape improvement by rounding . . . 86
4.22 Mixing volume possible layout . . . 86
5.1 Top view (a) and front view (b) sketches of parallel confined jet test section 90 5.2 Sketch of probe orientation and 12 scan locations. . . 91
5.3 Scan section coordinates and dimensions. . . 92
5.4 Computational domain used for Full Navier-stokes (FNS) simulation . . . . 93
5.5 Whole domain model . . . 94
5.6 Particular of the computational domain for runs from 2 to 6 . . . 95
5.7 Cross section grid . . . 96
5.8 Measured mean velocities, U and V . . . 98
5.9 Measured and computed velocity U . First run. . . . 100
5.10 Measured and computed velocity V . First run. . . . 101
5.11 Measured and computed velocity U . Second run. . . 102
5.12 Measured and computed velocity V . Second run. . . 103
5.13 Velocity U . Second and fifth run. . . 104
5.14 Velocity V . Second and fifth run. . . 105
5.15 Measured and computed cross flow velocities. First run. . . 106
5.16 Measured and computed cross flow velocities. Second run. . . 107
5.17 Cross flow velocities. Second run and fifth run. . . 107
5.18 Measured axial velocity inlet profile . . . 108
5.19 Run 2: axial velocity contour plot in the pipe . . . 109
5.20 Comparison between measured and computed inlet axial velocity profile . . 110
5.21 Jet decay along the confinement . . . 111
5.22 Jet decay along the confinement . . . 112
5.23 Measured and computed velocity U . Sixth run. . . 114
5.24 Measured and computed velocity V . Sixth run. . . 115
5.25 Measured and computed cross flow velocities. Sixth run. . . 116
5.26 Measured and computed axial velocity inlet profile . . . 117
5.27 Jet decay along the confinement . . . 118
6.1 IRIS Vessel. In the red circle the downcomer is shown. . . 120
6.2 Steam generator outlet cross section . . . 121
6.3 Modeled portion of the downcomer . . . 121
6.4 Modeled domain cross section. In the red circles barrel and vessel walls. . . 122
6.5 Steam generator outlet cross section. . . 123
6.6 Computational domain . . . 125
6.7 Particular of the mesh: transition layers . . . 127
6.8 Traces of the axial velocity profiles. . . 129
6.10 Axial velocity profiles in the long inlet pipe. . . 130
6.11 Velocity contour plot: radial plane . . . 131
6.12 Velocity contour plot: circumferential plane . . . 132
6.13 Recirculation zones: radial plane . . . 133
6.14 Recirculation zones: circumferential plane . . . 134
6.15 Entrainment in the jet of the surrounding water inventory . . . 135
6.16 In blue: halfjet line. In red: center line. In brown: nextjet line. . . 137
6.17 Full scale model: axial velocity . . . 138
6.18 Residence time down scaled: axial velocity . . . 140
6.19 Plug inlet flow profile and near wall treatment . . . 141
List of Tables
2.1 Reactor Vessel parameters . . . 9
2.2 Steam generators parameters . . . 14
2.3 Implications of safety-by-design approach . . . 16
2.4 IRIS response to PWR Class IV events . . . 17
2.5 IRIS electricity generation target costs . . . 20
3.1 Three level synthesis . . . 49
3.2 FSA at component level . . . 49
3.3 FSA at system level . . . 50
4.1 PIRT for downcomer and lower plenum during normal operation . . . 57
5.1 Comparison between IRIS and Kunz facility dimensionless parameters . . . 91
5.2 Mesh parameters for the first run . . . 95
5.3 Mesh parameters for the runs 2-5 . . . 96
5.4 Fluid properties, volume flow rates and mean velocities . . . 97
6.1 Computational domain dimensions . . . 120
6.2 Computational domain length along the flow direction . . . 124
6.3 Mesh parameters . . . 126
6.4 Fluid properties and flow conditions in the full scale simulations . . . 127
6.5 Fluid Properties and flow conditions in down scaled simulations . . . 128
Chapter 1
Introduction
Phenomena that need to be considered in the nuclear power plant design are very complex and their analysis requires computational analysis by sophisticated tools, since full scale experimentation is possible only in a limited number of cases. One-dimensional system codes have been used for a long time in this purpose. However, the flow in components such as the lower plenum and the downcomer of the Reactor Pressure Vessel (RPV) is clearly 3D. Natural circulation and mixing in containment volumes are also clearly 3D phenomena and in some analysis this characteristic can not be neglected. The availability of powerful computers and of efficient numerical techniques can heavily influence the development of utilized tools. Particularly, thermo-hydraulic analysis of nuclear power plants can now take into account the utilisation of CFD. Even if CFD for single phase flow has reached such a reliability that it can be used to solve a great number of problems of industrial interest, in the nuclear industry, because of the safety issues, a further effort in CFD model validation is needed.
Large experimental databases are needed in code assessment. Unfortunately an in-novative design, like the IRIS reactor design, can exploit only a part of the available experimental data and with a limited reliability. In this context, as a member of the IRIS consortium, the University of Pisa (Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione) has the task to build a down scaled experimental facility. Designing and running the experimental facility need awareness about involved phenomena and a proper scaling strategy.
It must to be observed that plant and/or component behavior is not equally influenced by all processes and phenomena that occur within them. An optimum analysis reduces candidate phenomena to a manageable set by identifying and ranking the phenomena with respect to their influence on figures of merit. Cause and effect are also differentiated. After the processes and phenomena have been identified, their importance should be determined with respect to their effect on the relevant figures of merit. The principal product of the
2
process illustrated above is a Phenomena Identification and Ranking Table (PIRT). Later on, CFD can confirm and quantify this preliminary analysis.
CFD, even when a full validation is not yet completed, can also provide data to ad-dress a proper scaling analysis. The scaling analysis should be conducted to ensure, by demonstrating the relevance and sufficiency of the collective experimental database, that the data, and the models based on them, will be applicable to the full scale analysis of the plant.
The purpose of the scaling analysis is to provide:
1. the design parameters for reduced-size test facilities; 2. the conditions for operating experiments;
3. the non-dimensional parameters which facilitate the efficient and compact presen-tation and correlation of experimental results;
4. the quantitative criteria to substantiate or revise the ranking of phenomena; 5. the basis for quantifying scale distorsions;
6. the scaling criteria for simulating component interactions within a system.
The present thesis work is aimed at applying CFD tools in the validation of a PIRT developed for the lower plenum and downcomer phenomena in the IRIS reactor. In fact, an analysis of thermo-hydraulic phenomena occurring in the downcomer and lower plenum region of the IRIS reactor were completed. One of the identified phenomenon has been chosen for a further CFD analysis. Before simulating the entrance of the coolant water in the downcomer, the CFD tool has been verified on a similar experimental case.
