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. Each phase consisted of several exercises. It was desirable also to plan the benchmark specification to accept as many potential numerical approaches as possible.
The Nuclear Power Engineering Corporation (NUPEC) Pressurised Water Reactor (PWR) Sub-channel and Bundle Tests (PSBT) Benchmark consisted of two phases, and each phase contained three exercises. Phase I was the Void Distribution Benchmark and Phase II was the Departure from Nucleate Boiling (DNB) Benchmark. Participants could choose either of the two phases and any of the exercises within the phases to take part and contribute. The preliminary indications showed that a sufficient number of participants attempted both phases with different numerical approaches.
In addition to the measured experimental data and the relevant boundary conditions, the detailed geometrical data of mock-up assemblies, spacers and the test loop were included as far as possible in the specification to allow a wide range of numerical modelling. Below are the corresponding exercises of the two phases.
The goal of this exercise was to benchmark the sub-channel, mesoscopic and microscopic numerical approaches. The experimental data included computed tomography (CT) scan measurements of the sub-channel averaged void fraction in four sub-channel types: typical central, central with a guide tube, side and corner. Also, graphical images of the void distribution within the typical central sub-channels and central sub-channel with guide tube were made available. The test cases were selected at PWE-rated conditions. Different types of single sub-channel test assemblies were used to investigate the effect of geometry on the phenomenon concerned.
This exercise was designed for benchmarking mesoscopic numerical approaches. The experimental data includes X-ray densitometer measurements of the void fraction (chordal averaged over the four central sub-channels) at three axial elevations along the bundle length and graphical images of the bundle void distribution. The test cases for this exercise were chosen at PWR rated conditions and deviations of quality from the rated conditions.
The NUPEC PSBT Database included simulation of four representative transients of PWRs; power increase, flow reduction, depressurisation and temperature increase. All four transients were selected as benchmark cases. The experimental data included time histories of X-ray densitometer measurements of the void fraction (chordal averaged over the four central sub-channels) at three axial elevations along the bundle length for four transient scenarios: power increase, flow reduction, depressurisation and temperature increase. Exercise three of Phase I was designed for benchmarking sub-channel numerical approaches.
This exercise was designed for performing code-to-code comparisons concerning axial pressure drop. Although no empirical data was available, code results were to be compared with relevant graphical data.
The exercise was designed to assess the thermal-hydraulic code capabilities of predicting the exit coolant temperature. The experimental data included measurements of the sub-channel-averaged fluid outlet temperatures.
The goal was to assess the thermal-hydraulic code capabilities of correct prediction of DNB along rod bundles. The experimental data includes the power at which DNB occurs and the corresponding locations in the bundle.
The exercise was designed to enhance the development of truly mechanistic models for DNB prediction during the four postulated transients in PWRs. The experimental data included the time histories of the boundary conditions and the DNB time detected for four transient scenarios: power increase, flow reduction, depressurisation and temperature increase.
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Data related to PSBT can be requested from the NEA Data Bank:
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 the topic is currently addressed in the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS).
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.
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.