Thermal hydraulic analysis of ETRR-2 using RELAP5 code
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I. D. Abdelrazek
Abstract
The present work was developed within the frame of the IAEA Coordinated Research Project 1496 “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”. Three benchmark experiments were designed and performed to study the performance of the Egyptian Research Reactor ETRR-2. The experiments included steady state measurements as well as loss of flow transient (LOFT) and loss of power transient (SCRAM) conditions. The Code RELAP5/Mod3.4 was used to simulate the components of the ETRR-2 systems for the thermal hydraulic analysis of the reactor. The dimensions and elevations of the primary cooling components are based on real conditions (3-D configuration). RELAP5 results provided benchmark data which verified the experimental measurements taken from instrumentations installed for this experiment at several positions in the core. The predicted values for the coolant and clad surface temperature at different locations in the core showed a remarkable agreement with the experimental values, for both the steady state and transient conditions.
Kurzfassung
Im Rahmen eines von der IAEA koordinierten Forschungsprojektes wurde das Betriebs- und Störfallverhalten des ägyptischen Forschungsreaktors ETRR-2 untersucht. Dazu wurden die Komponenten des ETRR-2 mit den Modellen des Programms RELAP5/Mod 3.4 abgebildet. Rechnungen zur stationären Betriebsverhalten und zu den Ereignissen loss of flow and loss of power wurden experimentellen Daten gegenübergestellt. Dabei zeigte sich bei der Bestimmung verschiedener Kerntemperaturen eine gute Übereinstimmung.
References
1 International Atomic Energy Agency: Proceeding Series on “RR Benchmarking Database: facility description and experiments”, ETRR-2 Nuclear Reactor (Egypt), Facility specification, Vienna (2012)Suche in Google Scholar
2 RELAP5/MOD3 Code Manual, RELAP5/MOD3.2, NUREG/CR-5535, INEL-95/0174Suche in Google Scholar
3 International Atomic Energy Agency: Proceeding Series on “RR Benchmarking Database: facility description and experiments”, ETRR-2 Nuclear Reactor (Egypt), Steady state experiment description, Vienna (2012)Suche in Google Scholar
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5 International Atomic Energy Agency: Proceeding Series on “RR Benchmarking Database: facility description and experiments”, ETRR-2 Nuclear Reactor (Egypt), LOFA experiment, Vienna (2012)Suche in Google Scholar
6 International Atomic Energy Agency: Proceeding Series on “RR Benchmarking Database: facility description and experiments”, ETRR-2 Nuclear Reactor (Egypt), Reactivity insertion – Experiment description, Vienna (2012)Suche in Google Scholar
© 2015, Carl Hanser Verlag, München
Artikel in diesem Heft
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- Summaries/Kurzfassungen
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- Tarapur atomic power station: analysis of station blackout scenario
- Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD
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- Transient performance during stopping the research reactor primary loop pump
- Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method
- A study on the criticality of modified neutron transport equation by using alternative scattering phase functions
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Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluation of advanced displacement cross-sections for the major EUROFER constituents based on an atomistic modelling approach
- Tarapur atomic power station: analysis of station blackout scenario
- Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD
- Thermal hydraulic analysis of ETRR-2 using RELAP5 code
- Transient performance during stopping the research reactor primary loop pump
- Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method
- A study on the criticality of modified neutron transport equation by using alternative scattering phase functions
- Production cross–section calculations of medical 32P, 117Sn, 153Sm and 186,188Re radionuclides used in bone pain palliation treatment
- Study on carbonated hydroxyapatite as a thermoluminescence dosimeter
- Social impact theory based modeling for security analysis in the nuclear fuel cycle
- Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor
