Under the guidance of the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS), the Expert Group on Reactor Core Thermal-hydraulics and Mechanics will perform specific tasks associated with multi-scale core thermal-hydraulics and mechanics of present and future nuclear power systems with focus on multi-phase flow heat transfer mechanisms. Coupling aspects of fluid structural mechanical interactions for normal and/or deformed core configurations will be considered, as will problems requiring high fidelity multi-scale core simulation. The considered reactor systems for which core thermo-hydraulic and structural mechanical issues will be studied include but are not limited to:
The ojective is to provide expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for multi-scale core thermal-hydraulics modelling and simulation of existing and proposed nuclear reactor systems.
A key activity associated with this objective is the identification and preservation of appropriate experimental data. In addition, the EGTHM will provide member countries with guidance and processes for certifying the experimental data for its use as a stand-alone core thermal-hydraulic validation or coupled thermal-hydraulics and structural mechanics. The role of thermal-hydraulics as part of validation pyramid of multi-physics modelling and simulation tools will be done in close collaboration with the Expert Group on Reactor System MUlti-Physics (EGMUP).
EGTHM will provide specific technical information regarding:
The expert group will monitor, steer and support the continued development of the TIETHYS database. The expert group also will facilitate the dissemination of technical information and knowledge through activities such as workshops, benchmark studies and training activities.
Alessandro Petruzzi (ITA)
Maria N. Avramova (USA)
All NEA member countries EGTHM members' working area
|European Commission (under the NEA Statute)
Observer (international organisation)
|International Atomic Energy Agency (by agreement)
WPRS and associated Expert Group meetings are held approximately every twelve months.
Click on elements in graph above to jump directly to a specific WPRS or EGTHM activity.
Lead-cooled Fast Reactors (LFR) are rather new concepts which have gathered increasing international attention after the Generation-IV International Forum (GIF) selected them as promising candidates for a new generation of nuclear energy systems. The LRF physics benchmark is based on the Advanced LFR European Demonstrator (ALFRED) design and consists of a neutronics and thermal hydraulics stage with each three benchmark phases related to pin-cell, sub assembly/super-cell and whole-core simulations.
The Liquid Metal Fast Reactor (LFMR) is one of the next generation reactor designs. This benchmark consists of steady-state numerical predictions of Texas A&M University (TAMU) separate effect tests and of numerical predictions of the Thermal Hydraulic Out of Reactor Safety (THORS) integral effect tests and comparison to experimental results.
High-temperature gas-cooled nuclear reactors are one of the next generation reactor designs. This benchmark consists of a multi-staged code-to-code-to-data thermal hydraulics code validation benchmark using data measured at the High Temperature Test Facility (HTTF) at Oregon State University (OSU).