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). The CANDU Owners Group (COG) provided full-scale experimental data for a 28-element CANDU fuel assembly including pressure drop and detailed pin temperatures for single-phase and boiling conditions.
The purpose of the benchmark is to:
A unique aspect of these data is the traversing thermocouples which provide axial and azimuthal sheath temperatures along every fuel element for both pre= and post-CHF conditions. A wide range of test conditions (inlet pressure, mass flows and inlet temperatures) will be provided to participants along with the temperature scans, CHF location and power, and post-dryout sheath temperatures. Similar to the NUPEC Boiling Water Reactor (BWR) Full-size Fine-mesh Bundle Tests (BFBT) and NUPEC Pressurised Water Reactor (PWR) Subchannel and Bundle Tests (PSBT) benchmarks, the complete set of data will be made available to participants after completion of the benchmark.
NEA: O. Buss
McMaster University Co-ordinator: D.R. Novog
Created in 2019, under the guidance of the Working Party on Scientific Issues of Reactor Systems (WPRS) the expert group will perform specific tasks associated with core thermal-hydraulics aspects of present and future nuclear power systems.
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.
The Working Group on the Analysis and Management of Accidents (WGAMA) is responsible for activities related to potential accidental situations in nuclear power plants, including the following technical areas: reactor coolant system thermal-hydraulics; design-basis accidents; pre-core melt conditions and progression of accidents and in-vessel phenomena; coolability of over-heated cores; ex-vessel corium interaction with coolant and structures; in-containment combustible gas generation, distribution and potential combustion; physical-chemical behaviour of radioactive species in the primary circuit and the containment; and source term. The activities mainly focus on existing reactors, but also have application for some advanced reactor designs. Priority setting is based on established CSNI criteria and in particular on safety significance and risk and uncertainty considerations.
The Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS) studies the reactor physics, fuel performance, and radiation transport and shielding in present and future nuclear power systems.