Emergency Core Cooling Injection Research Articles

Risk modelling of ageing nuclear reactor systems

A nuclear reactor is expected to function for extensive periods, during which, coolant circulation and core reactivity must always be maintained safely. Understanding the risks associated with the operation of such systems requires proper consideration of ageing components and the effects of preventative maintenance. The traditional methodologies, such as Fault Trees and Event Trees, have limitations in their abilities to model ageing processes and complex maintenance strategies. Petri Nets have been used in this research as a more suitable alternative. A case study reactor is presented to demonstrate this capability. Petri Nets were developed for five key subsystems: primary coolant circulation, shutdown condensation, emergency core coolant injection, emergency shutdown, and control and monitoring, building a representation which considers their failure modes, reaction of the system to faults, and ongoing component maintenance actions. These models reveal statistics for the timing of failure of these subsystems and relative frequencies of outcome categories.

Computational studies on pressurized thermal shock in reactor pressure vessel

Pressurized Thermal Shock (PTS) in the Reactor Pressure Vessel (RPV) can occurs during injection of Emergency Core Cooling (ECC) water into the cold leg of Pressurized Water Reactor. Injected ECC water generates large temperature gradient which leads to large thermal stresses in the RPV wall. To study this behaviour in Rossendorf Coolant Mixing Model (ROCOM), case of buoyancy driven mixing with a constant flow rate and 10% density difference between the ECC injection and loop water was taken for analysis. ROCOM is a 1:5 scaled German KONVOI type reactor having four-loop with a RPV mock-up made of transparent acrylic material. Large Eddy Simulation (LES) and Reynolds Averaged Navier–Stokes equations (RANS) based models are used for this numerical analysis work in OpenFOAM CFD code. LES offered more accurate predictions of the local and spatial mixing phenomena compared to RANS based models. The predicted results are compared and validated with experimental data of ROCOM test facility. It shows that results are in good agreement.

Numerical prediction of a single phase Pressurized Thermal Shock scenario for crack assessment in an Reactor Pressure Vessel wall

A pressurized thermal shock (PTS) in a reactor pressure vessel (RPV) wall of a Pressurized Water Reactor (PWR) could eventually lead to reduced integrity of the RPV. A PTS scenario may be initiated by a Loss-of-Coolant Accident (LOCA) in which the Emergency Core Cooling (ECC) injection causes a thermal shock in the vessel wall while the RPV is still partially pressurized. The traditional Thermal-Hydraulic system codes fail to reliably predict the complex three-dimensional thermal mixing phenomena in the downcomer occurring during the ECC injection. Therefore, a three-dimensional computational fluid dynamics (CFD) simulation of the transient is required to provide an accurate temperature distribution for reliable probabilistic analyses. The assessment of the failure probability due to a single phase PTS initiated by postulated defects in the vessel wall requires the computation of a long transient solution. Since the CFD calculation is computationally expensive, the symmetry of a four-leg PWR geometry is utilized to reduce the computational domain.In the present work, several Unsteady Reynolds Averaged Navier-Stokes (URANS) CFD simulations are performed to select an optimal computational domain that is representative for a typical PWR. The computed temperature distribution is provided as input for the structural analyses to calculate the Stress Intensity Factor (SIF) along several postulated defects. The SIF is then used to evaluate the potential of crack propagation initiation of the defects.

IAEA CRP benchmark of ROCOM PTS test case for the use of CFD in reactor design using the CFD-Codes ANSYS CFX and TrioCFD

Over the last 15 years, considerable effort has been expended in assembling the available information on the use of CFD in the nuclear reactor safety field. Typical application areas here are heterogeneous mixing and heat transfer in complex geometries, buoyancy-induced natural and mixed convection, etc., with specific reference to Nuclear Reactor Safety (NRS) accident scenarios such as Pressurized Thermal Shock (PTS), boron dilution, hydrogen build-up in containments, thermal fatigue and thermal striping issues, etc. The development, verification and validation of CFD codes in respect to Nuclear Power Plant (NPP) design necessitates further work on the complex physical modelling processes involved, and on the development of efficient numerical schemes needed to solve the basic equations. Therefore, a set of ROCOM CFD-grade test data were made available to set up an International Atomic Energy Agency (IAEA) benchmark, relating to PTS scenarios. The benchmark deals with the injection of the relatively cold Emergency Core Cooling (ECC) water, which can induce buoyancy-driven stratification. Data obtained from the PTS experiment were compared in the study presented here with predictions obtained from the CFD software packages ANSYS CFX and TrioCFD. In addition a test case without buoyancy forces was selected to show the influence of density differences.The results of the two test cases and the numerical calculations show that mixing efficiency is strongly influenced by buoyancy effects. At higher mass flow rates without density differences the injected slug propagates in the circumferential direction around the core barrel. Buoyancy forces reduce this azimuthal propagation. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor 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 the observed flow patterns and mixing phenomena.

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Abstract A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the Primarkreislauf-Versuchsanlage (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

Design of a single-phase PTS numerical experiment for a reference Direct Numerical Simulation

The integrity assessment of the Reactor Pressure Vessel (RPV) is considered to be an important issue for lifetime extension of nuclear reactors. A severe transient that can threaten the integrity of the RPV is the existence of a Pressurized Thermal Shock (PTS) during a Loss-of-Coolant Accident (LOCA). A PTS consists of a rapid cooling of the RPV wall under pressurized conditions that may induce the criticality of existing or postulated defects inside the vessel wall. The most severe PTS event has been identified by Emergency Core Cooling (ECC) injection during a LOCA. The traditional one-dimensional system codes fail to reliably predict the complex three-dimensional thermal mixing phenomena in the downcomer occurring during the ECC injection. Hence, CFD can bring real benefits in terms of more realistic and more predictive capabilities. However, to gain trust in the application of CFD modelling for PTS, a comprehensive validation programme is necessary. In the absence of detailed experimental data for the RPV cooling during ECC injection, high fidelity Direct Numerical Simulation (DNS) databases constitute a valid alternative and can serve as a reference.The aim of this work is to design a numerical experiment aimed to generate a high quality reference DNS database for a simplified PTS scenario. This takes into account the turbulent mixing in the downcomer and the evolution of the temperature distribution for both structures and fluid during a single-phase flow PTS scenario. In spite of simplifications, such a DNS analysis represents a very demanding application. A priori, it should be demonstrated that all the relevant turbulent scales will be fully resolved, which requires a huge computational power. A wide range of Reynolds Averaged Navier–Stokes (RANS) calculations are performed in order to determine a realistic PTS computational domain. Furthermore, the resulting numerical experiment is optimized in terms of geometry, boundary conditions, and physical properties.

ADVANCED DVI+

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.

Application of large-eddy simulation to pressurized thermal shock problem: A grid resolution study

Abstract Life time extension assessment is a very important issue for the nuclear community. A serious threat to the life time extension of a Reactor Pressure Vessel (RPV) is an occurrence of the Pressurized Thermal Shock (PTS) during an Emergency Core Coolant (ECC) injection in a loss-of-coolant accident. Traditional system codes fail to predict the complex three-dimensional flow phenomena resulting from such ECC injection. Improved results have been obtained using Computational Fluid Dynamics (CFD) analysis based on the Reynolds-Averaged Navier–Stokes (RANS) equations. However, it has been also shown that transient RANS approaches are less capable to predict the complex flow features in the downcomer of the RPV. More advanced CFD methods like Large-Eddy Simulation (LES) are required for modeling of these flow phenomena. This paper addresses the feasibility of the application of LES for single-phase PTS. Furthermore, the required grid resolution for such LES analyses is identified by evaluation of the solution on different mesh sizes. A buoyancy-driven PTS experiment has been considered here. This experiment has been performed in the 1:5 linear scale Rossendorf Coolant Mixing Model (ROCOM) facility. In the applied numerical model, the incompressible Navier–Stokes equations are solved in the LES formulation, with an additional transport equation for a scalar, which is responsible for driving the embedded buoyancy term in the momentum equations. Instantaneous mixing characteristics are investigated based on evaluation of the scalar concentration. The analysis presented in this paper indicates that the application of LES is feasible nowadays for single-phase PTS. It is demonstrated that the mixing in the downcomer is quite sensitive to small turbulent disturbances at the ECC inlet, i.e., two simulations performed with slightly different fluctuations at the inlet result in substantially different flow in the downcomer. This complicates the analysis of the data from simulations and suggests that evaluation of the results should be performed in the frequency–amplitude domain instead of the classically employed temporal data analysis.

Main results of the European project NURESIM on the CFD-modelling of two-phase Pressurized Thermal Shock (PTS)

Abstract The European Platform for NUclear REactor SIMulations, (NURESIM project 2005 – 2008) addressed the creation of a Common European Standard Software Platform for modelling, recording, and recovering computer data for nuclear reactors simulations. One work package of the project was dedicated to the analysis and improvement of CFD capabilities for the simulation of two-phase PTS problems. Some SB-LOCA scenarios lead to a situation in which the cold leg is partially or totally uncovered when the Emergency Core Cooling injection is activated. The resulting complex two phase flow can be divided in characteristic flow regions: the jet flow with a free surface between steam and water, the zone of jet impingement, the horizontal two-phase flow and the flow in the downcomer. Many phenomena have to be reflected in a simulation of each separate region, but also when the simulations are coupled reflecting the integral process which is required to predict the thermal loads at the RPV wall. After analyzing the experimental database available for CFD model development and validation and identifying shortcomings of the models different activities were dedicated to the simulation of single flow regions as well as the integral flow. Based on these experiences recommendations for the CFD-simulation of the two-phase PTS problem were obtained.

Analysis of Low-Pressure ECC Injection Bypass of an ABWR During a Feedwater Line Break with RELAP5-3D/K

In the innovative design of the advanced boiling water reactor (ABWR), conventional recirculation loops are removed and replaced by multiple reactor internal pumps. Therefore, there is no major penetration of the reactor pressure vessel (RPV) below the elevation of the top of active fuel. As a result, an ABWR loss-of-coolant accident (LOCA) can have a decreased impact on reactor safety. Moreover, in the new RPV design the injection points of all the conventional low-pressure emergency core cooling (ECC) systems (ECCSs) are shifted out of the core shroud to the downcomer and feedwater line as a new low-pressure ECCS, namely, a low-pressure flooder (LPFL). Consequently, the net hydraulic head built inside the downcomer will be the only driving force to bring the low-pressure ECC water into the core shroud during a large-break LOCA. In the analysis of a feedwater line break with RELAP5-3D/K, it was occasionally found that the hydraulic head built in the downcomer might not be great enough to bring the ECC water into the core shroud, and when the mixture water column ascends above the elevation of the feedwater rings, all the water injected by the LPFL will be directly driven to the break on the feedwater line. Fortunately, the capacity of the remaining high-pressure ECC flow directly injected above the core is great enough, and this ECC low-pressure injection bypass phenomenon can be terminated once the high-pressure ECC injection is manually turned off. This phenomenon of low-pressure ECC injection bypass is unexpected in the ABWR design, and it is worth further investigation.

An Overview of the Pressurized Thermal Shock Issue in the Context of the NURESIM Project

Within the European Integrated Project NURESIM, the simulation of PTS is investigated. Some accident scenarios for Pressurized Water Reactors may cause Emergency Core Coolant injection into the cold leg leading to PTS situations. They imply the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. This paper focuses on two-phase conditions that are potentially at the origin of PTS. It summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor,that is, the cold leg and the downcomer, where the “PTS fluid-dynamics" is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two-phase flow subphenomena are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance is still limited. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet well controlled, leading to slightly conservative predictions.

Experimental and Numerical Modeling of Transition Matrix From Momentum to Buoyancy-Driven Flow in a Pressurized Water Reactor

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the cold leg and downcomer of a pressurized water reactor (PWR) was analyzed at the Rossendorf coolant mixing (ROCOM) test facility. This paper presents a matrix of ROCOM experiments in which water with the same or higher density was injected into a cold leg of the reactor model with already established natural circulation conditions at different low mass flow rates. Wire-mesh sensors measuring the concentration of a tracer in the injected water were installed in the cold leg, upper and lower part of the downcomer. A transition matrix from momentum to buoyancy-driven flow experiments was selected for validation of the computational fluid dynamics software ANSYS CFX. A hybrid mesh with 4×106 elements was used for the calculations. The turbulence models usually applied in such cases assume that turbulence is isotropic, whilst buoyancy actually induces anisotropy. Thus, in this paper, higher order turbulence models have been developed and implemented, which take into account that anisotropy. Buoyancy generated source and dissipation terms were proposed and introduced into the balance equations for the turbulent kinetic energy. The results of the experiments and of the numerical calculations show that mixing strongly depends on 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 circumferential propagation with lower mass flow rates and/or higher density differences. 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 the ECC injection in PWR and should be also considered in pressurized thermal shock scenarios. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

CFD Code Validation against Stratified Air-Water Flow Experimental Data

Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

MODELING OF A BUOYANCY-DRIVEN FLOW EXPERIMENT IN PRESSURIZED WATER REACTORS USING CFD-METHODS

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 Pressurised 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. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. 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 circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Three-dimensional numerical calculations have been performed for a transient moderator circulation inside the CANDU (Canada Deuterium Uranium) calandria vessel of Wolsong Units 2/3/4. The porous media approach was applied for the core region containing 380 calandria tubes. An anisotropic hydraulic resistance model for the porous media has been developed based on the empirical pressure loss correlations. The selected event was the 35% RIH (Reactor Inlet Header) break with a loss of ECC (Emergency Core Cooling) injection, which has been known to give the largest heat load to the moderator among all the DBA's (Design Basis Accidents). The calculation has been successfully done until 1,200 s after the break, when most of the considerable heat transfer procedure has been completed. During this LOCA (Loss of Coolant Accident) transient, the local subcoolings in the vicinity of any PT/CT (Pressure Tube/Calandria Tube) contact does not drop below the experimentally derived subcooling threshold of 30°C. Because the minimum subcoolings reach only a few degrees to the threshold temperature during the initial 20-40 s, future work on the CANDU moderator circulation needs to be aimed at determining whether this small subcooling margin covers the uncertainty of the moderator analysis.

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.

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Embrittlement behavior of zircaloy-4 cladding during oxidation and water quench

Simulated loss of coolant accident (LOCA) tests on Zircaloy-4 cladding were carried out to evaluate the thermal shock property during the injection of emergency core coolant. A Zircaloy-4 specimen was oxidized in a steam environment between 1000 and 1250°C followed by a flooding of the cooling water. After the test, the ductility of the thermally embrittled specimen was measured by a ring-compression test and a microstructural analysis was carried out. The results showed that the threshold equivalent cladding reacted (ECR) value to cause a failure was higher than the conventional 17% criterion calculated by the Baker–Just equation. A residual metal thickness under 0.3mm as well as a ring-compression ductility below 0.2mm for a fracture is effective to assess the thermal shock embrittlement of Zircaloy-4 in an axially unrestrained, or even in a restrained condition.

Effect of the yaw injection angle on the ECC bypass in comparison with the horizontal DVI

The comparison tests for the direct emergency core cooling (ECC) bypass fraction were experimentally performed with a typical direct vessel injection (DVI) nozzle and an ECC column nozzle having a yaw injection angle to the gravity axis. The ECC yaw injection nozzle is newly introduced to make an ECC water column in the downcomer region. The yaw injection angle of the ECC water relative to the gravity axis is varied from 0 to (±)90° stepped by 45°. The tests are performed in the air–water separate effect test facility (direct injection visualization and analysis (DIVA)), which is a 1/7.07 linearly scaled-down model of the APR1400 nuclear reactor. The test results show that (1) if the ECC water column is injected into the wake region which is induced by the hot leg blunt body in the downcomer annulus, the ECC bypass fraction is greatly reduced compared with the typical horizontal ECC injection which makes ECC film on the downcomer wall. At the same time, the ECC penetration toward the lower downcomer region becomes larger than those of a typical horizontal type of direct vessel injection on the downcomer wall vertically. (2) If the ECC water column is injected near the broken cold leg, the ECC water is directly bypassed. Thus, the ECC penetration fraction is greatly reduced compared with a typical film type of the horizontal ECC injection. (3) In order to minimize the ECC bypass fraction, the ECC water should be injected toward the wake region of the hot leg blunt bodies.

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments.