Impact of cross-section generation procedures on the simulation of the VVER-1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3D thermal-hydraulics/neutron kinetics models

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 a l'Enerige Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark is to assess computer codes used in 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). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3D thermal-hydraulics/neutron kinetics models using the data available in the benchmark specifications and a common cross-section library. The first code-to-code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between two sets of results. The present paper focuses on the analysis of the observed discrepancies. The VVER-1000 3D neutron kinetics models are based on cross-section data homogenized on the assembly level. The cross-section library, provided as part of the benchmark specifications, thus consists in a set of parameterized two group cross sections representing the different assemblies and the reflectors. The origin of the observed large discrepancies was found mainly to lie in the methods used to solve the diffusion equation. The VVER reflector properties were also found to enhance discrepancies by increasing flux gradients at the core/reflector interface thus highlighting more the difficulties in some codes to handle high exponential flux gradients. This paper summarizes the different steps applied to analyze the neutronic codes and their predictions as well as the impact of cross-section generation procedures.     

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).     

Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.     

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.     

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.     

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.     

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.     

Validation of coupled codes using VVER plant measurements

A data set of five transients at different VVER type nuclear power plants was collected in order to validate neutron kinetics/thermal hydraulics codes. Two of these transients ‘drop of control rod at nominal power at Bohunice-3’ of VVER-440 type and ‘coast-down of 1 from 3 working MCPs at Kozloduy-6’ of VVER-1000 type, were then utilised for code validation. Eight institutes contributed to the validation with 10 calculations using 5 different combinations of coupled codes. The thermal hydraulic codes were ATHLET, SMABRE and RELAP5 and the neutron kinetic codes DYN3D, HEXTRAN, KIKO3D and BIPR8. The general behaviour of both the transients was quite well calculated with all the codes. Even an elementary modelling of coolant mixing in reactor pressure vessel under asymmetric transients improved correspondence to the measurements. Some differences between the calculations seem to indicate that fuel modelling and treatment of VVER-440 control rods need further consideration. The simultaneous validation interacted with the data collection effort and thus improved its quality. The complexity of data collection systems and sometimes conflicting data, however, called for compromises and interpretation guides that also taught the analysts balanced plant modelling.     

Comprehensive uncertainty and sensitivity analysis for coupled code calculations of VVER plant transients

The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics codes, is an important step to perform best-estimate calculations for plant transients of nuclear power plants. For applications in safety analysis, these coupled codes should be validated by benchmark calculations and, preferably, by comparison with plant transient data from operating plants. In addition, the results should be supplemented by applying uncertainty and sensitivity analysis methods, which allow to identify relevant parameters of models and solution procedures affecting the results and to quantify their relative importance. Both objectives were part of the VALCO project. The aspect of validation is presented in [S. Mittag, et al., 2004. Neutron-Kinetic Code Validation against Measurements in the Moscow V-1000 Zero-Power Facility, in press; T. Vanttola et al., 2004. Validation of coupled codes using VVER plant measurements, in press], the aspect of a comprehensive uncertainty and sensitivity analysis for coupled code calculations is the topic of this contribution. The results and experiences obtained by the analysis for two plant transients in a VVER-440 and a VVER-1000, respectively, are presented and discussed.     

Validation of coupled neutron kinetic/thermal-hydraulic codes. Part 2: Analysis of a VVER-440 transient (Loviisa-1)

Several three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal-hydraulic system codes. Under the auspices of the European Union's Phare programme these codes have been validated against real plant transients by the participants from 7 countries. Two of the collected five transients were chosen for validation of the codes. Part 1 of this article consists of validation against VVER-1000 reactor data. This second part is focussed to validation against measured data of ‘One turbo-generator load drop experiment' at the Loviisa-1 VVER-440 reactor. The experiment was performed just after plant modernisation and more measured data was available to validation than in normal operation of real plants. Good accuracy of the results was generally achieved comparable to the measurement accuracy. The confidence in the results of the different code systems has increased, and consequences of certain model changes could be evaluated.     

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.     

Effect of steam environment on severe core damage behaviour for VVER-1000 with the ASTEC V1 code

Severe accident studies for very low frequency events for VVER-1000 (V320) are carried out to estimate in-vessel damage progression under steam-rich and starved conditions. The analyses with code ASTEC, jointly developed by IRSN (France) and GRS, Germany), have shown the influence of steam environment on core heat-up followed by material relocation, hydrogen production, vessel failure and aerosol generation along with release to containment. Hydro-accumulator injection for studied transients also gives rise to a steam-rich environment enhancing the material oxidation depending on the injection time and period. The generated information along with PSA-Level 2 is helpful to decide Plant Damage State (PDS) and fruitfully develop accident management strategies for the plant.     

A spatial kinetic model for simulating VVER-1000 start-up transient

An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means “Reactor spatial kinetic”. All required group constants in calculations are prepared using the WIMS code. In addition CITATION code was used to calculate the flux, power distribution and core reactivity inside the core. To update the last change in group constants and resultant reactivity in point kinetic equations, these neutronic codes were coupled with a developed dynamic program. This study is applied on a typical VVER-1000 reactor core to show the reactor response in short time transients caused during start-up procedure.     

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.     

Neutron-kinetic code validation against measurements in the Moscow V-1000 zero-power facility

Measurements carried out in an original-size VVER-1000 mock-up (V-1000 facility, Kurchatov Institute, Moscow) were used for the validation of three-dimensional neutron-kinetic codes, designed for VVER safety calculations. The significant neutron flux tilt measured in the V-1000 core, which is caused only by radial-reflector asymmetries, was successfully modeled. A good agreement between calculated and measured steady-state powers has been achieved, for relative assembly powers and inner-assembly pin power distributions. Calculated effective multiplication factors exceed unity in all cases. The time behaviour of local powers, measured during two transients that were initiated by control rod moving in a slightly super-critical core, has been well simulated by the neutron-kinetic codes.     

Numerical boron mixing studies for Loviisa nuclear power plant

A program is in the process of studying numerically boron mixing in the downcomer of Loviisa NPP (VVER-440). Mixing during the transport of a diluted slug from the loop to the core might serve as an inherent protection mechanism against severe reactivity accidents in inhomogenous boron dilution scenarios for PWRs. The commercial general purpose Computational Fluid Dynamics (CFD) code PHOENICS is used for solving the governing fluid flow equations in the downcomer geometry of VVER-440. So far numerical analyses have been performed for steady state operation conditions and two different pump driven transients. The steady state analyses focused on model development and validation against existing experimental data. The two pump driven transient scenarios reported are based on slug transport during the start of the sixth and first loop, respectively. The results from the two transients show that mixing is case and plant specific; the high and open downcomer geometry of VVER-440 seems to be advantageous from mixing point of view. In addition the analyzing work for the ‘first pump start' scenario brought up some considerations about flow distribution in the existing experimental facilities.     

AFEN method and its solutions of the hexagonal three-dimensional VVER-1000 benchmark problem

The unique features of the analytic function expansion nodal (AFEN) method in hexagonal-z geometry are described. The COREDAX code implementing the AFEN method is verified testing on the VVER-440 benchmark problem and a “simplified” VVER-1000 benchmark problem. The COREDAX code then applied to the original VVER-1000 benchmark problem exercise 2 (HZP case and HP case) provides very good results in comparison with those of other benchmark participants.     

Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project “TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.     

Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO

Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.