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.     

High-resolution boron dilution measurements using laser induced fluorescence (LIF)

The application of the laser induced fluorescence technique to the study of liquid mixing in the downcomer of a pressurized water reactor is presented. The scenario is that of a boron dilution event, in which a deborated slug is set in motion by the actuation of a reactor coolant pump. A separate effects test facility, built with transparent plexiglas, is used to conduct optical measurements of the slug mixing along its path to the core. The optical assembly is described and the conditions for the implementation of laser induced fluorescence as a quantitative measurement technique are discussed. Results from a slug injection experiment are shown which demonstrate the high-resolution capabilities of this procedure as applied to the study of liquid mixing in the complex geometry of a reactor vessel downcomer.     

Boron mixing in complex geometries: flow structure details

Quantitative measurements of scalar dilution transients within a scaled model of a PWR downcomer are presented. The injection transients correspond to the case of a single pump start-up injecting into an initially stagnant system. Detailed results of the flow path and velocity field are presented for a nominal case where a neutrally buoyant front of dilute water is injected into the downcomer. Additional tests are made for a finite-sized slug of dilute water with a volume approximately equal to that of the downcomer and the lower plenum. The results of the tests indicate that the flow is dominated near the injection leg by an impinging jet, which is bifurcated into two primary streams by the cylindrical geometry of the downcomer. The two streams exit the downcomer on the opposite side of the injection leg. The splitting of the streams creates a bell-shaped recirculation region in the lower downcomer beneath the injection leg, which is the last volume to be displaced by the injected fluid flowing up from the lower plenum. A third plume sometimes forms at the boundary between the streams and the recirculation region, introducing mixed fluid into parts of this relatively stagnant area at earlier times than is typically observed. The results are discussed in the context of computational fluid dynamics codes that are starting to be used to simulate the mixing phenomena encountered in such complex systems.     

Fluid mixing and flow distribution in the reactor circuit, measurement data base

Experimental investigations and computational fluid dynamics (CFD) calculations on coolant mixing in pressurised water reactors (PWR) have been performed within the EC project FLOMIX-R. The project aims 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. Measurement data from a set of mixing experiments have been gained by using advanced measurement techniques with enhanced resolution in time and space. Slug mixing tests simulating the start-up of the first main circulation pump are performed with two 1:5 scaled facilities: the Rossendorf Coolant Mixing model ROCOM and the Vattenfall test facility. Additional data on slug mixing in a VVER-1000 type reactor have been gained at a 1:5 scaled metal mock-up at EDO Gidropress. Experimental results on buoyancy driven mixing of fluids with density differences have been obtained at ROCOM and the Fortum PTS test facility.Concerning mixing phenomena of interest for operational issues and thermal fatigue, flow distribution data available from commissioning tests at PWRs and VVER are used together with the data from the ROCOM facility as a basis for the flow distribution studies.In the paper, the experiments performed are described, results of the mixing experiments are shown and discussed. Efforts on computational fluid dynamics codes validation on selected mixing tests applying Best Practice Guidelines in code validation will be reported about in a separate paper.     

CFD simulation of Fortum PTS experiment

Detailed simulation of the thermal stresses of the reactor pressure vessel (RPV) wall in case of pressurized thermal shock (PTS) requires the simulation of the thermal mixing of cold high-pressure safety injection (HPI) water injected to the cold leg and flowing further to the downcomer. The simulation of the complex mixing phenomena including, e.g., stratification in the cold leg and buoyancy driven plume in the downcomer is a great challenge for CFD methods and requires careful validation of the used modelling methods.The selected experiment of Fortum mixing test facility modelling the Loviisa VVER-440 NPP has been used for the validation of CFD methods for thermal mixing phenomena related to PTS. The experimental data includes local temperature values measured in the cold leg and downcomer. Conclusions have been made on the applicability of used CFD method to thermal mixing simulations in case with stratification in the cold leg and buoyant plume in the downcomer.     

Ex-vessel transport and mixing of a deborated slug in a PWR primary geometry

The transport and mixing of a slug of deborated water in a lowered loop PWR is modeled by partitioning the volumes of the primary system according to chemical rector theory. Piping is modeled as plug flow volumes while the steam generator outlet plenum and the reactor coolant pumps are modeled as backmixed volumes. This simple approach provides a good representation of the transport and mixing phenomena outside the reactor vessel. The proposed methodology can be used to generate initial and boundary conditions for separate effects tests and CFD computations for the reactor vessel complex geometry. The decoupling of the ex-vessel primary system greatly enhances the resolution of boron dilution transient issue.     

Comparative study of a boron dilution scenario in VVER reactors

Subsequent studies have identified many scenarios, which can lead to reactivity excursions due to boron dilution. The comparative study, presented in this paper, deals with the so-called “restart of the first reactor coolant pump’’ scenario and its reactor-dynamic consequences for both Russian designed VVER reactor types, VVER-440 and VVER-1000. The transient simulations were performed using the three-dimensional core dynamics code DYN3D. The DYN3D modeling features, including recent developments, as well as the cross-section methodology involved in these calculations, are described. The analyzed accident scenario is outlined together with the assumptions made. The results of core response in this boron dilution accident for both VVER reactors are compared within the ranges, determined by the two reactivity values of interest: the criticality limit and the reactivity initiated accident (RIA) limit.     

Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD codes CFX-5 and Trio_U

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the downcomer of a pressurized water reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields.An experiment with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water was selected for validation of the CFD software packages CFX-5 and Trio_U. Two similar meshes with approximately 2 million control volumes were used for the calculations. The effects of turbulence on the mean flow were modeled with a Reynolds stress turbulence model in CFX-5 and a LES approach in Trio_U. CFX-5 is a commercial code package offered from ANSYS Inc. and Trio_U is a CFD tool which is developed by the CEA-Grenoble, France.The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: at higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this propagation. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. Both CFD codes were able to predict well the observed flow patterns and mixing phenomena.     

Numerical simulation of boron injection in a BWR

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark’s Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several minutes.     

First experience with a six-loop nodalization of a VVER-440 using a new coupled neutronic-thermohydraulics system KIKO3D-RETINA V1.1D

In this paper, we introduce a new, coupled neutronic-thermohydraulics system. The three-dimensional neutron kinetic code KIKO3D and the two-phase flow code RETINA V1.1D have been coupled for modeling complex transients of nuclear power plants. Using a six-loop nodalization of a VVER-440, several test calculations have been carried out. Results obtained for a trip of one main circulation pump are compared with real measurements and reference calculations provided by other neutronic-thermohydraulics systems. The ability of our coupled system is demonstrated.     

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.     

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.     

SBLOCA with boron dilution in pressurized water reactors. Impact on operation and safety

When the water level in the reactor pressure vessel (RPV) of a pressurized water reactor (PWR) is low enough and the core temperature is such that the coolant in that region boils, reflux-condensation conditions are established. Under such conditions, almost boron-free water is collected in a region of the primary system forming a non-borated slug. If subsequent natural circulation is established or a reactor coolant pump (RCP) is restarted, the slug could be transported to the core. This scenario configures an important part of the so-called boron issue. The Energy Systems Analysis Group at the Institute of Energy Technologies (INTE) of the Technical University of Catalonia (UPC) has studied the boron issue in three different stages. The steps were the following: participation in OECD-related projects, code improvement and investigation at nuclear power plant (NPP) scenarios. The third step is the main aim of this paper and consists of a continuation of the previous projects in the field of NPP analysis. The aim of this paper is to study SBLOCA transients with boron dilution in PWR. The chosen NPP was Ascó-2 which is a 3-loop-2940,6 MWth Westinghouse PWR. The paper contains some references to OECD/SETH and OECD/PKL experimental projects and analyses an established scenario including features of boron transport and sensitivity calculations for relevant parameters.     

Core response of a PWR to a slug of under-borated water

A parameter study, incorporating stationary and transient core calculations, was carried out for a hypothetical boron dilution event in a pressurized water reactor, using the three-dimensional (3D) neutron kinetics core model DYN3D in combination with a fast running semi-analytical coolant mixing model. It was assumed that a slug of deborated water was formed in one of the loops, due to a secondary-to-primary steam generator leakage during outage. It was further assumed that this slug is not recognised and that the first main coolant pump is started, in preparation to returning to power. At the initiation of pump start-up, the reactor is still in the cold and deeply sub-critical state. By varying the initial slug volume, it was found in stationary calculations that, for the given core configuration, slugs of less than 14 m³ do not lead to re-criticality. Transient core calculations with larger slug volumes show a significant reactivity insertion and over-criticality. However, according to the calculations, even an over-criticality of about 2$ did not lead to safety-relevant consequences. The power excursion is mitigated and stopped by Doppler feedback. The influence of the cross-section library on the course of the transient was investigated, applying an alternative library. Differences in the global core parameters lead to quantitative differences in the time and height of the power excursion. In addition, it is shown that numerical diffusion has to be suppressed in order to describe the boron transport correctly, especially at low flow velocities. Otherwise the neutron kinetic core behaviour cannot be correctly modelled.     

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.     

Two-phase flow in a full-scale loop seal facility

An experimental program has been carried out to study two-phase behaviour of a PWR cold leg loop seal during loss-of-coolant accidents. The experimental facility comprises a full-scale cold leg with a reactor coolant pump simulator. Three separate air/water test series were performed to determine the onset of slugging in the horizontal pipe, the residual water mass and the total two-phase pressure drop in the loop seal.The results of flow regime transition experiments have been compared with smaller-scale experiments and with theoretical predictions to evaluate scaling criteria. The strong hysteresis of transitions found between the stratified and slug flow regimes depends on the loop seal geometry and U-tube oscillations.     

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.     

Validation of the CFD code fluent by post-test calculation of a density-driven ROCOM experiment

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems.During the course of follower core assessments, TÜV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core inlet has to be determined.TÜV NORD SysTec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events, a validation against the ROCOM experiment T6655_21 with a density-driven coolant mixing was performed. This validation proves that FLUENT is able to appropriately simulate the effects of boron transport and dilution such as streaks of coolant with lower density in the downcomer. Deficits were identified in the simulation of fluid layering in the cold leg, which fortunately have a rather small influence on the predicted core inlet concentration. Therefore, the boron concentration in the reactor core can be determined with sufficient accuracy to solve the safety issue, regardless of the core becoming critical or not.     

RELAP5/MOD3.2 investigation of a VVER-1000 MCP switching on Benchmark problem

This paper provides comparisons between experimental data of Kozloduy NPP “MCP switching on when the other three MCP are in operation”, with Relap5 calculations. The investigated thermal-hydraulic driven transient is characterized by spatially dependant non-symmetric processes. 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. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. The main purpose of this investigation was to improve the discrepancy between the calculations and the plant data. The sensitivity calculation investigates the mixing in reactor vessel and influence of heat structure on the hot legs temperature. The areas of improvements to the Relap5 model are:

• •The non-symmetrical mixing in downcomer and reactor vessel annular exit.
• •The influence of heat structure temperature on the time delay for equipments measurements.
• •Investigation of pressurizer water level – using the hot legs temperature correction.

The RELAP5/MOD3.2 model of Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios have been developed and validated in the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS) Sofia, and Kozloduy NPP. The model provides a significant analytical capability for the specialists working in the field of NPP safety.This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons indicate good agreement between the RELAP5 results and the experimental data. The sensitivity investigation improves the discrepancy between the calculation and the plant data.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 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.