Expert Group on Uncertainty Analysis in Modelling (EGUAM)

Completed

The expert group provided advice to the Working Party on Scientific Issues of Reactor Systems (WPRS) on the scientific development needs (data and methods, validation experiments, scenario studies) of sensitivity and uncertainty methodology for modelling of different reactor systems and scenarios. It was completed in 2020 and its scope of work is currently addressed in the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS).

Related topics

Nuclear science

Reactor physics

Multiphysics

Publications and reports

results

SEE ALL IN THIS CATEGORY

          
        NEA Document
      
	        Most recent

OECD/NRC Benchmark Based on NUPEC Pressurised Water Reactor Sub-channel and Bundle Tests (PSBT) - Volume I: Experimental database and final problem specifications

4 January 2012

Benchmark EGTHM English Multiphysics Multiphysics NSC Nuclear science …

          
        Publications and Reports
      
	        Most recent

NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark -  Volume II: Uncertainty and Sensitivity Analyses of Void Distribution and Critical Power - Specification

1 January 2010

Benchmark EGTHM English Light water reactors Modelling and simulation Nuclear science Nuclear science …

          
        NEA Document
      
	        Most recent

International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests - Volume III: Departure from nucleate boiling

2 October 2015

Benchmark EGTHM English Multiphysics NSC Nuclear science Nuclear science …

Objectives

The main activity was focused on uncertainties in modelling light-water reactor (LWR) transients. In this context the objectives were to:

determine modelling uncertainties for reactor systems under steady-state and transient conditions, quantifying the impact of uncertainties for each type of calculation in the multi-physics analysis, i.e.:
- neutronics calculations
- thermal-hydraulics modelling
- fuel behaviour
determine for each of the above types of calculation the major sources of uncertainty arising from:
- data (e.g. nuclear data, geometry, materials)
- numerical methods
- physical models
develop and test methods for combining the above sources of uncertainty for each type of calculation so as to yield uncertainty assessment for the coupled multi-physics analyses.
develop a benchmark framework which combined information from available integral facility and nuclear power plant experimental data with analytical and numerical benchmarking:
- where available, experimental data was used to test the individual types of calculation as well coupled multi-physics simulations

Members

EGUAM members' area (password protected | reminder)

Contact

wprs@oecd-nea.org

Activities

results

Benchmark Based on NUPEC Pressurised Water Reactor (PWR) Subchannel and Bundle Tests (PSBT)

An approach similar to the OECD/NRC Boiling Water Reactor Full-size Fine-mesh Bundle Tests (BFBT) Benchmark was adopted by dividing the benchmark activity into phases.

Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM)

The goal of the Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM) is to determine the uncertainty in light water reactor (LWR) systems and processes in all stages of calculations. It is estimated through a simulation process of ten exercises in three phases provided by the benchmarking framework.

Benchmark of the Modular High-Temperature Gas-Cooled Reactor-350 MW Core Design

Helium-cooled very high-temperature gas reactors are highlighted as a key technology with the potential to improve the competitiveness of nuclear energy within the Generation IV International Forum (GIF). Developing tools and methods to support this technology is seen as a priority by NEA member countries.

See

Boiling Water Reactor Turbine Trip (BWRTT) Benchmark

The Boiling Water Reactor Turbine Trip (BWRTT) Benchmark was established to challenge the coupled system thermal-hydraulic/neutron kinetics codes against a Peach-Bottom-2 (a GE-designed BWR/4) turbine trip transient with a sudden closure of the turbine stop valve.

See

Expert Group on Physics of Reactor Systems (EGPRS)

Under the guidance of the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS), the Expert Group on Physics of Reactor Systems (EGPRS) performs specific tasks associated with reactor physics aspects of present and future nuclear power systems.

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Expert Group on Reactor Systems Multi-Physics (EGMUP)

The Expert Group on Reactor Systems Multi-Physics (EGMUP) advances the state of the art in establishing processes and procedures for certifying experimental data and benchmarking multi-physics multi-scale modelling and simulation (M&S).

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KALININ-3 Coolant Transient Benchmark

This benchmark was a continuation of the V1000CT activities and defined a coupled code problem for further validation of thermal-hydraulics system codes for application with Russian-designed VVER-1000 reactors based on actual plant data from the Russian nuclear power plant Kalinin Unit 3 (Kalinin-3)

McMaster Core Thermal-Hydraulics Benchmark

Building on the NEA BFBT and PSBT benchmarks, McMaster University, in collaboration with North Carolina State University, launched a CANDU thermal-hydraulic benchmark under the WPRS Expert Group on Uncertainty Analysis in Modelling (EGUAM).

OECD/NRC Benchmark based on NUPEC BWR Full-size Fine-mesh Bundle Tests (BFBT)

NUPEC Boiling Water Reactor (BWR) Full-size Fine-mesh Bundle Test (BFBT) Benchmark

Pebble Bed Modular Reactor (PBMR) Coupled Neutronics/Thermal-hydraulics Transients Benchmark - PBMR-400 Core Design

The international benchmark concerned pebble bed modular reactor (PBMR) coupled neutronics/thermal-hydraulics transients based on the PBMR-400MW design.

Physics of plutonium recycling

For nuclear utility companies that opt to reprocess spent fuel, recycle it or do both, one of the main pertinent issues is plutonium recycling.

See

Pressurised Water Reactor Main Steam-Line Break (MSLB) Benchmark

This benchmark incorporated full 3-D modelling of the reactor core into system transient codes for best-estimate simulations of the interactions between reactor core behaviour and plant dynamics and their testing on a number of transients of importance for plant behaviour and safety analysis.

Pressurised Water Reactor MOX/UO2 Core Transient Benchmark

PWR MOX/UO2 Core Transient Benchmark

Subgroup on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM)

The Subgroup on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM) was formed to check the use of best-estimate codes and data.

VVER-1000 Coolant Transient (V1000CT) Benchmarks

The overall objective of the VVER-1000 Coolant Transient (V1000CT) Benchmark was to assess computer codes used in the safety analysis of water-water energetic reactor (VVER) power plants, specifically for their use in reactivity transients in VVER-1000.