Rostov-2 VVER-1000 Benchmark
Ongoing

Introduction

The Rostov-2 benchmark tests the capabilities of 3D multi-physics reactor simulation codes to analyse complex transients with coupled core-plant interactions and complicated fluid mixing phenomena. The benchmark activity aims to compare the performance of traditional multi-physics codes and novel high fidelity multi-physics code systems and methods, and to provide recommendations on advantages and disadvantages of traditional and novel approaches.

Experiments

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.

Benchmark Description

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 [1], 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.

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Benchmark Activities on Reactor Single- and Multi-Physics of the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS)

The Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS)  and its expert groups co-ordinate benchmark activities on Reactor Single- and Multi-Physics. The benchmark activities serve the WPRS objectives to: 

  • provide Members with up-to-date information, preserve knowledge on, and develop consensus regarding reactor physics, thermal-hydraulics, radiation transport and dosimetry, and multi-physics aspects associated with nuclear power systems, aimed at providing the technical underpinnings of assessment of system performance and safety.
  • provide advice to the nuclear community on the developments needed to meet the requirements (data and methods, validation experiments, scenario studies) for the assessment of different reactor systemsprovide training and educational material to the nuclear community, to assist with developing future technical expertise within member countries. Workshops will be held on WPRS benchmarks
  • provide training and educational material to the nuclear community, to assist with developing future technical expertise within member countries. 

 The activities address the the following reactor types:

  • Current fleet of light-water and heavy water reactors (LWRs/HWRs) as well as present generation of fuel designs.
  • Evolutionary and innovative LWRs/HWRs along with advanced and/or accident tolerant fuel designs.
  • Next generation of reactor systems including water cooled small modular reactors (SMRs), micro-reactors, high temperature gas cooled reactors (HTGR) as well as advanced fast spectrum systems including sodium fast reactors (SFRs) and molten salt reactor systems (MSRs).
  • Accelerator driven (sub-critical) and critical systems for waste transmutation as well as fusion systems.