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 benchmark team selected a test (transient), which allows validation of novel high-fidelity multi-physics codes developed during last years in the frame of different national and international projects (e. g. NURESAFE in European Union (EU), as well as Consortium for Advanced Simulation of LWRs (CASL), and Nuclear Energy Advanced Modeling and Simulation (NEAMS) in United States of America (USA), etc.). The difference in comparison with all previous multi-physics OECD/NEA benchmarks for coupled code validation is the implementation of high fidelity multi-physics simulation codes that could predict pin-by-pin power distributions and flow mixing in the reactor pressure vessel including its active core part.
The Rostov-2 VVER-1000 benchmark has two phases with a suite of sequential exercises. The first phase is designed to provide the framework to assess the ability of the traditional multi-physics (coupled system thermal-hydraulic (TH)/neutronics (N)) codes to predict the transient response of the power plant on assembly wise level. That means that assembly wise homogenisation of parameters for the thermal-hydraulics and neutron physics will be applied.
Phase I Exercise 1 - Point kinetics thermal hydraulics plant (system) simulation
The purpose of this exercise is to test the primary and secondary system models’ responses. Compatible point kinetics model input data (feedback coefficients, CR 10 differential reactivity in a form of tables, averaged axial power distribution, etc.) obtained using a 3D code neutronics model of the core or utilizing directly some measured data are provided in electronic format.
Phase I Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulic response evaluation
The purpose of this exercise is to model the reactor pressure vessel with the active core only. Inlet and outlet core transient boundary conditions are provided by the benchmark team based on calculations performed with coupled system code or applying any information directly from the measured data. Calculation of the hot zero power (HZP) state of the core (Exercise I-2a) is a part of this exercise to test the neutronics model including the cross-section library. Exercise I-2b includes calculations of initial hot power (HP) steady-state conditions and the transient test scenario simulation. The required two-group assembly-wise homogenized cross-section library is provided in electronic format. In case of interest the participants can generate their own two-group assembly-wise homogenized cross-section library.
Phase I Exercise 3 - Best-estimate coupled code plant transient modeling
This exercise combines elements of the first two exercises in this benchmark and is an analysis of the transient in its entirety. For participants that have already taken part in the Kalinin-3 Benchmark , it is suggested to start directly with this exercise. As a preface step for these participants is recommended to perform steady state core calculations at HZP state and HP initial state and deliver results for comparisons. That will ensure a check for the correct application of the cross-section libraries, the core loading and the core design geometry. Each participant should compare his results with the local data (SPND at 7 axial levels and temperature data located at assemblies’ heads) and with global/integral data (reactor power, temperature in hot and cold legs, etc.) of measured parameters.
The second phase of the benchmark is designed to provide a framework to assess the ability of the novel multi-physics codes to predict the transient response of the NPP on the base of pin-by-pin level. For this purpose, new developments that allow performing of high fidelity thermal-hydraulics and neutronics calculations will be utilised. Such codes could be coupled TH system codes with TH sub-channel codes (multi-scale thermal-hydraulic system-core coupling). The thermal-hydraulic core model is further coupled with improved fuel models and neutronic pin-by-pin homogenised models or pin-by-pin heterogeneous models. Pin-by-pin power reconstruction procedures could also be used for this purpose if other methods are not available. Methods and codes that can perform partially core assembly wise coupled thermal-hydraulics/neutronics calculations and partially pin-by-pin coupled thermal-hydraulic/neutronics even at the level of one hot channel/assembly pin-by-pin calculations are welcomed. The specifications of this phase are still in development.
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)
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.