VVER-1000 Coolant Transient Benchmark Research Articles

CFD analysis of coolant mixing in VVER-1000/V320 reactor pressure vessel

This study presents a code-to-code and model-to-model comparison of coolant mixing in the VVER-1000/V320 Kozloduy Unit 6 nuclear power plant using Computational Fluid Dynamics (CFD). Four different CFD codes were used to simulate coolant mixing in the reactor vessel, namely ANSYS Fluent, ANSYS CFX, TrioCFD, and STAR-CCM+. Two different approaches were used to model the upper plenum, while a single simplified model was used for the reactor pressure vessel. The simulations were performed for VVER-1000 coolant transient benchmark (V1000CT-2) mixing exercise. The results were compared between the different CFD codes and models to assess the accuracy and consistency of the simulations with the available experimental data. Overall, the results showed good agreement between the different CFD codes and models, with minor differences observed in some cases. The simplified models were found to be sufficient for predicting the overall coolant mixing patterns observed in the reactor vessel, provided additional insights into the local flow structures and mixing characteristics. This study demonstrates the applicability and reliability of CFD simulations for coolant mixing analysis in VVER-1000/V320 nuclear power plants.

Validation of the Serpent 2-HEXTRAN-SMABRE code sequence in a VVER-1000 coolant transient

ABSTRACT This paper presents a VVER-1000 coolant transient benchmark calculated using Serpent 2 generated group constants in the reactor dynamics code HEXTRAN coupled with the system code SMABRE. Two different sets of group constants applying different leakage correction methods are generated. The purpose of the paper is to validate the Serpent 2-HEXTRAN–SMABRE code sequence in plant transient calculations for a commercial VVER-1000 reactor. Calculated radial power distributions, relative fission power, mass flow rate and coolant heat-up are compared to experimental data. These factors, axial power distributions, radial and axial peaking factors and axial offset are also compared to results calculated using HELIOS generated group constants provided in the benchmark. Results calculated with Serpent generated group constants agree with the experimental data mostly within measurement uncertainty. Deviations larger than measurement uncertainty are explained by the insufficiency of available data needed for the HEXTRAN–SMABRE simulation or the albedo calculations.

Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

The Institute of Neutron Physics and Reactor Technology (INR) is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV) as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT) and include asymmetric vessel flow transients and main steam line break (MSLB) transients. Plant data from flow mixing experiments are available for comparison. Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

Application of coulped code method for the analysis of VVER-1000 coolant transient benchmark

In the light of the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations, which were not possible few years ago, can now be performed. Nowadays, it becomes possible to switch to new generation of computational tools, namely, coupled code (CC) technique. The application of such method is mandatory for the analysis of transient events where strong coupling between the core neutronics and the primary circuit thermal-hydraulics exits, and more especially when asymmetrical processes take place in the core leading to local space-dependent power generation. Through the current study, a demonstration of the maturity level achieved in the calculation of 3-D core performance during complex accident scenarios is emphasized. The study is followed by a typical application through which the main features and limitations of this technique are discussed.

Validation Cases of CATHARE 2 for VVER-1000 Main Steam Line Break Analysis

Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plant transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation.

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.

Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

ICONE15-10327 VVER SAFETY ANALYSIS BY THE COUPLED CODE ATHLET/BIPR-VVER

The paper describes the modelling features for a VVER-1000 nuclear power plant by the coupled system code ATHLET/BIPR-VVER. Special attention is paid on the reactor pressure vessel model and its nodalization schema. They play an important role for the realistic prediction of the fluid mixing phenomena in case of asymmetric transients. For such a transient - simultaneous isolation of one steam generator from the steam line and the feed water supply at low power level - validation to measured parameters are presented and analysed. The results of the simulation are compared with experimental local and integral plant loop data from NPP Kozloduy Unit 6 collected during the plant-commissioning phase. Additional examples of coupled code application for VVER-1000 reactor applying detailed reactor pressure vessel nodalization are also presented, choosing Exercise 3 of the CEA-NEA/OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) - a main steam line break.

HEXTRAN–SMABRE calculation of the VVER-1000 coolant transient-1 benchmark

The VVER-1000 coolant transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three-dimensional neutron kinetic core models. It concerns switching on a main coolant pump when the other three main coolant pumps are in operation. The problem is based on an experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, an extreme scenario concerning a control rod ejection after switching on a main coolant pump was calculated. At VTT the three-dimensional advanced nodal code HEXTRAN is used for the core dynamics, and the system code SMABRE as a thermal hydraulic model for the primary and secondary loop. The parallelly coupled HEXTRAN–SMABRE code has been in production use since early 1990s, and it has been extensively used for analyses of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used at VTT. The whole core calculation is performed with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation were specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Parametric studies have been performed for selected parameters.

Comparison of RELAP5 calculations of VVER-1000 coolant transient benchmark phase 1 at different power

This paper provides comparisons between experimental data of “MCP switching on when the other three MCPs are in operation” and RELAP5 calculations with different initial levels of the reactor power 29.45% and 27.47% from the nominal. The reference power plant for this analysis is Unit 6 at the Kozloduy nuclear power plant (NPP) site. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

Simulation of the VVER-1000 pump start-up experiment in the OECD/DOE/CEA V1000CT benchmark by the FLICA4/CRONOS2 coupled code system

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat à l'Energie Atomique (CEA), France, a coupled 3-D thermal–hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark [Ivanov, B., Ivanov, K., Groudev, P., Pavlova, M., Hadjiev, V., 2002. VVER-1000 Coolant Transient Benchmark (V1000-CT). Phase 1 – Final Specifications, NEA/NSC/DOC] is to assess computer codes used in the analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The CEA presented results for the V1000CT-1 Exercise 2 using a coupling of FLICA4 [Toumi, I., Gallo, D., Bergeron, A., Royer, E., Caruge, D., 2000. FLICA4: a three dimensional two-phase flow computer code with advanced numerical methods for nuclear applications. Nuclear Engineering and Design 200, 139–155] and CRONOS2 [Akherraz, B., Baudron, A.M., Lautard, J.J., Magnaud, C., Moreau, F., Schneider, D., Gonzales, M., 2004. Manuel de Référence CRONOS 2.6. Technical Report SERMA/LENR/RT/04-3433/A, CEA] via the coupling tool ISAS [Toumi, I., et al., 1995. Specifications of the general software architecture for code integration in ISAS. Euratom Fusion Technology, ITER task S81TT-01/1]. The FLICA4/CRONOS2 VVER-1000 model is based on the data available in the benchmark specifications. This paper summarizes the FLICA4/CRONOS2 model build-up with the associated sensitivity studies and presents the CEA results for V1000CT-1 Exercise 2 as well as a comparison with experimental results at hot power steady state (HP SS).

NRI Řež solution of Exercise 3 of VVER-1000 Coolant Transient Benchmark – Phase 1 with coupled codes DYN3D–ATHLET

The paper presents a solution of VVER-1000 Coolant Transient Benchmark – Phase 1 (V1000CT-1) of Exercise 3 performed with the coupled reactor dynamic code DYN3D and system code ATHLET at NRI Řež. The first part of the paper contains brief characteristics of VVER-1000 NPP input deck and describes also the applied reactor core model. The second part introduces the steady-state results and important time dependencies, compared with experimental values. The calculation results show that such type of transient can be realistically described by the coupled codes DYN3D–ATHLET.

Modelling effects and their impact on the results of ATHLET solutions of Exercise 1 of the VVER-1000 Coolant Transient Benchmark

The VVER-1000 Coolant Transient Benchmark consists of two phases and refers to experimental data from the Kozloduy Unit 6 Nuclear Power Plant in Bulgaria. The paper describes the modelling features and their impact on the results of the Exercise 1, Phase 1 of the Benchmark obtained by two ATHLET user groups, namely GRS and NRI. The simulated transient is a main coolant pump (MCP) switching on in one loop at reduced power while three other MCPs are in operation. Particular attention is paid to the influence of the reactor vessel modelling and especially of the nodalization in the upper plenum. The comparison and discussion of the two simulation results confirm that the two solutions with the ATHLET system code achieve quite good system response of the plant transient.

OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark – A consistent approach for assessing coupled codes for RIA analysis

The rod ejection accident (REA) and the main steam line break (MSLB) are two of the most important design basis accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal-hydraulics codes for reactivity insertion accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that to perform cross code comparative analysis for REA and MSLB scenarios. The coupled 3-D neutron kinetics/thermal-hydraulics benchmark presented in this paper is based on data from the Unit #6 of the Bulgarian Kozloduy Nuclear Power Plant (KNPP) and it is entitled the VVER-1000 coolant transient (V1000CT) benchmark. Two real plant transients are selected for simulation in the benchmark: main coolant pump start-up (Phase 1) and coolant mixing tests (Phase 2). In addition to these transients extreme scenarios were defined for better testing 3-D neutronics/thermal-hydraulics coupling: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP (Phase 1) and MSLB transient (Phase 2). The paper presents an overview of the Phase 1 (V1000CT-1) benchmark activities and describes the approach used for assessing the coupled neutron kinetics/thermal-hydraulics codes. Selected comparative analysis of currently submitted participants' results is presented with emphasis on the observed modeling issues and deviations from the measured data.

Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation. The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results.

Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled system code RELAP5/PARCS

As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 coolant transient benchmark phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PARCS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant, Unit 6, and the neutron kinetics models of PARCS for hexagonal geometries. Selected results of performed investigations will be presented and discussed in this paper. The overall trends of most plant parameters are in a reasonable agreement with the experimental data. Nevertheless multidimensional thermal hydraulic models are needed for a more realistic description of the coolant mixing phenomena within the reactor pressure vessel.