Hydraulic behaviour of a partially uncovered core

Directorate-General For Translation European Commission

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Nuclear energy and safety

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• Publié par: EUROPEAN-COMMISSION
• Nombre de lectures: 14
• Langue: English
• Poids de l'ouvrage: 2 Mo

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fuR. il.Hob
Commission of the European Communities
Nuclear Science and Technology
Shared Cost Action
Reactor Safety Programme 1985-1987
Hydraulic Behaviour of a partially Uncovered
Core
Final Report
Directorate-General for Science Research and Development
Joint Research Centre - Ispra Site
October 1989 EUR 12406 EN Commission of the European Communities
Nuclear Science and Technology
Shared Cost Action
Reactor Safety Programme 1985-1987
Hydraulic Behaviour of a partially Uncovered
Core
Final Report
Work performed in the frame of the
Shared Cost Action (SCA) programme 1985-87
Research Contract No. 3001-86-07 ELISPD
Author(s): K. Fischer and W. Hafner
Battel le - Institut e V.
Energy and Process Technologies Dept.
Am Romerhof 35
D-6000 Frankfort am Main
Directorate-General for Science Research and Developme
Joint Research Centre - Ispra Site I p . Bio'i *1
, EUR 12406 EN October 1989 i
C- I Published by the
COMMISSION OF THE EUROPEAN COMMUNITIES
Directorate-General
Telecommunications, Information Industries and Innovation
Batiment Jean Monnet
LUXEMBOURG
LEGAL NOTICE
Neither the Commission of the European Communities nor any person
acting on behalf of the Commission is responsible for the use which might
be made of the following information.
Cataloguing data can be found at the end of this publication.
Luxembourg: Office for Official Publications of the European Communities, 1989
ISBN 92-826-0770-4 Catalogue number: CD-NA-12406-EN-C
© ECSC - EEC - EAEC, Brussels-Luxembourg, 1989
Printed in Italy CONTENTS
1 1 INTRODUCTION
1.1 Loss—of—Coolant Accident (LOCA) Scenarios i
1.2 Aim and Objectives 2
1.3 Structure of the Report 3
2 THEORETICAL BACKGROUND 3
2.1 3 General Features of Code Simulation Techniques
2.2 3 Flow Regimes in the Partially Uncovered Core
2.3 5 Boundary Conditions for the Partially Uncovered Core
3 EXPERIMENTAL BACKGROUND 6
3.1 Description of Experiments 7
3.1.1 7 PERICLES
3.1.2 7 TLTA
7 3.1.3 ORNL
7 3.1.4 Westinghouse
7 3.1.5 FLECHT SEASET
3.1.6 ROSA-IV 8
3.1.7 ECN S
3.1.8 THETIS 8
3.1.9 Other Experiments 8
3.2 Geometrical Scaling O
9 3.3 Thermohydraulic Conditions
4 11 THERMOHYDRAULIC PROCESSES ABOVE THE SWELL LEVEL
4.1 Review of Experimental Data 11
4.1.1 12 PERICLES
4.1.2 TLTA 12
4.1.3 13 ORNL
4.1.4 13 Westinghouse
14 4.1.5 FLECHT SEASET
4.1.6 ROSA-IV 14
15 4.1.7 ECN and THETIS
4.1.8 Discussion of Experimental Data 15
4.2 Review of Theoretical Models 17 4.2.1 DRUFAN Models 19
4.2.2 COBRA/TRAC Models 20
4.2.3 RETRAN Models
4.2.4 TRAC-BD1 Models
4.2.5 THERMIT Models
4.2.6 TRAC-BWR Models1
4.2.7 TRAC-PF1 Models
4.2.8 RELAP 5 Model
4.2.9 Other Empirical Correlations
4.2.10 Droplet Interaction Models 22
4.2.11 Discussion of Theoretical Models3
4.3 Conclusions5
5 THERMOHYDRAULIC PROCESSES BELOW THE SWELL LEVEL 2
5.1 Review of Experimental Data6
5.1.1 PERICLES 2
5.1.2 TLTA
5.1.3 ORNL7
5.1.4 Other Experiments
5.1.5 Discussion of Experimental Data
5.2 Review of Theoretical Models
5.2.1 DRUFAN Models 31
5.2.2 COBRA/TRAC Models4
5.2.3 RETRAN Models
5.2.4 TRAC-BD1 Models8
5.2.5 THERMIT Models 4
5.2.6 LASL Models
5.2.7 TRAC-PF1 Models2
5.2.8 EPRI Drift Flux Model3
5.2.9 CATHARE Models6
5.2.10 THETIS Models7
5.2.11 RELAP5 Models9
5.2.12 ORNL Models 51
5.2.13 Discussion of Theoretical Models
5.3 Comparison ofl Models and Experimental Data 5
5.3.1 EPRI Full Range Correlation4
5.3.2 RELAP5/NEPTUN Interfacial Friction 5
5.3.3 GRS Void Correlations for the DRUFAN Code6 5.3.3.1 Viecenz-Mayinger Correlation (VM) 56
5.3.3.2 Wilson Correlation (WL)
5.3.3.3 Correlation for Slug Flow (SL) and Churn-turbulent Flow (CT) 5
5.3.3.4 Full Range Drift Correlation (FR)
5.3.4 CATHARE Interfacial Friction Model
5.4 Numerical Sensitivity Studies 60
5.4.1 Base Case
5.4.2 Variation of Interfacial Momentum Exchange1
5.4.3n of Pressure2
5.4.4n of Inlet Flow Rate4
5.4.5 Variation of Bundle Power6
5.4.6n of Axial Power Profile
5.5 Conclusions 67
6 ANALYTICAL STUDIES WITH COBRA-NC 68
6.1 ORNL Level Swell Test 3.09.10 AA9
6.2 FLECHT-SEASET Boil-Off Test 57 71
6.3 Discussion 75
7 CONCLUSIONS AND RECOMMENDATIONS6
8 REFERENCES8 1 INTRODUCTION
The thermohydraulic behaviour of light water reactor cooling systems under accident
conditions is simulated by large and complex computer codes. Especially, the core
coolability under loss-of—coolant accident conditions has to be demonstrated by numerical
simulations. The codes must be validated in comparison to experimental data in order to
make sure that code predictions are reliable. Integral system tests are performed on large
experimental facilities and simulated by the codes for to validate the overall performance of
the codes. Separate effects experiments and associated code simulations are done to develop
and validate the numerous models and submodels which are built into such a large system
code. In the present report, separate effects experiments and associated computer models for
the thermal—hydraulic processes in a partially uncovered core are reviewed.
1.1 Loss-of—Coolant Accident (LOCA) Scenarios
The large break LOCA is generally assumed to follow from the break of a main coolant
pipe in the primary loop of a light water reactor. The rapid pressure loss associated with the
blow—down leads to a complete core uncovering, followed by refill of the lower plenum and
reflood of the core under the action of the emergency core coolant injection systems. Times
until completion of core rewet are in the order of minutes. The thermal-hydraulic
conditions in the core are characterized by
— low pressures,
— high flow velocities,
— fast transients,
— core uncovering by steam flashing throughout the core,
— reflood by fast moving quench front.
The small break LOCA, on the other hand, is generally assumed to follow from special
system transients or operator actions. It is characterized by
— high or intermediate pressure,
— reduced flow velocities,
— slow transients,
— core dryout by boil—off,
— long—term partially uncovered core,
— slow motions of the mixture level.
An increased interest in the small break LOCA research came up after the TMI—2 accident
where a slow boil—off occurred about 100 minutes after the reactor scram, with a brief
flow reestablishment at 174 minutes. While the large break LOCA shows more drastic
changes in the system state, the small break LOCA may be more important to consider because of its higher probability of occurrence.
In the present work, the small break LOCA is considered as the governing event for
the partially uncovered core situation to be investigated. This means that special aspects of
the large break LOCA scenario like quench front propagation or counter-current flow
limitation are not considered.
1.2 Aim and Objectives
In the partially uncovered core, the upper uncovered region shows an increase in the rod
temperatures due to release of decay heat and insufficient heat removal by coolant flow. As
a first barrier against radionuclide release, the fuel rod claddings may get permeable for the
gases in the rods because of zircaloy oxidation. Further heating may lead to structural
degradation of rods with release of gases and aerosols from the fuel pellets to the primary
system. Reflooding the hot core terminates the thermal excursion, but it can introduce rod
damage due to thermal shocks. In summary, a wide spectrum of radionuclide releases to the
primary system can occur, depending on the amount and duration of the core heatup.
Another undesirable effect is the release of hydrogen from the zircaloy water reaction
to the primary system. Hydrogen in the primary system can deteriorate the core cooling.
When released to the containment atmosphere, hydrogen burn could lead to damage of
important components.
All these effects are strongly affected by the hydraulic processes inside the core. A
realistic model of the uncovered core hydrodynamics helps to determine the appropriate
system design and operator actions for the reduction of accident consequences and
radionuclide release. The present work is aimed at summarizing the experimental and
theoretical state—of—the—art of uncovered core thermal hydraulics and identifying necessary
additional research. Especially the development of the european codes DRUFAN/ATHLET
and CATHARE shall be supported. In detail, the following goals shall be achieved :
— set up a critical review of experimental data on partially uncovered core
thermohydraulics,
— set up a critical review of model approaches and modeling assumptions to simulate the
thermohydraulics of a partially uncovered core,
— perform sensitivity studies with a code to investigate parameter dependencies and
modeling shortcomings,
— specify requirements for future experiments to resolve modeling shortcomings, if
necessary.
The investigation is limited to intact rod structures; rod deformations due to overheating are
not considered.