The challenges constituted by the accurate and realistic simulation of some multi-physics phenomena are of great concern in the industrial environment, and the demand for an advanced reactor modelling tool of multiple physics phenomena has thus increased significantly in recent years. Pellet cladding interaction (PCI) has been identified as one of the more interesting multi-physics, multiscale problems since possible PCI fuel failures reduce reactor performance related to power uprates, higher burn-up and fuel rod manufacturing quality.
Under the guidance of the NEA Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV), the Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark was proposed by Nuclear and Industrial Engineering (N.IN.E.), in co-ordination with Studsvik, to:
The MPCMIV benchmark initiative is based on experiments that require coupling between reactor physics, thermal-hydraulics and fuel performance tools to achieve a high-fidelity simulation.
Selected cases involve cold ramp tests assessed in the Studsvik R2 tank-in-pool testing reactor (the R2 core domain) that has an in-pile U-Tube system loop where the fuel rodlet is positioned (the fuel rodlet domain). The fuel response was investigated at cold criticality conditions (below 100°C) since the cladding mechanical properties and the potential failure mechanisms could differ from those at normal operation.
The test was performed with an ad hoc proce-dure that first places the rod in cold conditions and then exposes it to a relatively fast transient, in which the maximum power generation in the rod increases from practically zero to 45 kW/m or more in about five seconds. The heat flux and fuel temperature reach their maximum values 10 to 15 seconds after the end of the power ramp, and the experiment ceases with a manual reactor shutdown.
A three-tiered structure of fidelity has been proposed so as to accommodate as many participants and computational tools as possible:
For each tier, the MPCMIV benchmark will be structured into four main phases:
Validation requirements will be set in all of the aforementioned steps.
The Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of Light Water Reactors (LWRs) is an international high-visibility benchmark for uncertainty analysis in best-estimate coupled code calculations for design, operation, and safety analysis of LWRs. The annual workshops are attended by many experts in industry, research institutes, national laboratories, academia, and government agencies.
The goal of the Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM) was to determine the uncertainty in light water reactor (LWR) systems and processes in all stages of calculations. It was estimated through a simulation process of nine exercises in three phases provided by the benchmarking framework.
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)
The aim of the benchmark is to improve understanding and modelling of pellet-cladding mechanical interaction (PCMI) amongst NEA member organisations. This is achieved by comparing PCMI predictions of different fuel performance codes for a number of cases.
A number of tests with detail well documented neutronics and thermal-hydraulics measurements data have been performed at the Rostov Unit 2 (Rostov-2) nuclear power plant (NPP). The reactor type is a VVER-1000 with fuel assemblies of type TBC-2M, which enable an 18-month fuel cycle length.
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.
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.