Emergency Core Cooling Research Articles

H212 Heat Pipe Based Emergency Core Cooling System for Nuclear Reactor

H212 Heat Pipe Based Emergency Core Cooling System for Nuclear Reactor

H213 二相流エジェクタによる無電源軽水炉緊急時炉心冷却系の理論解析(OS-11:沸騰・二相流(5))

H213 二相流エジェクタによる無電源軽水炉緊急時炉心冷却系の理論解析(OS-11:沸騰・二相流(5))

Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition

Pressure tube (zirconium 2.5 wt.% Nb) serves as a pressure boundary for the coolant that removes nuclear heat generated in the reactor core of Indian Pressurised Heavy Water Reactors (IPHWRs). Under postulated low frequency (<10 −6 per year) accidents like Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS) injection, heatup of pressure tube (PT) combined with internal pressure and the weight of the fuel bundle may lead to deformation. The extent and nature of deformation is important from reactor safety point of view. An experimental set-up has been designed and fabricated to simulate sagging (downward deformation) of PT due to its own weight and the weight of fuel bundles for 220 MWe IPHWRs. Experiments are conducted at different heatup rates of voided PTs. It is observed that sagging initiates at a temperature around 450 °C. Contact between PT and calandria tube (CT) occurs at around 585–625 °C, respectively. Once PT–CT contact takes place, PT temperature either decreases or the temperature rise remains controlled whereas CT temperature keeps on increasing for next 20–30 s. The contact location in all the experiments was near the centre of the tube. Structural integrity of PT is retained (no breach) for all the experiments. The PT temperature rise is found to be arrested after the contact between PT and CT, thus establishing that moderator acts as an efficient heat sink for IPHWRs.

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.

An Integral Effect Test on the Reflood Period of a Large-Break LOCA for the APR1400 Using ATLAS

A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), has been constructed at the Korea Atomic Energy Research Institute. It is a 1/2-reduced-height and 1/288-volume-scaled test facility based on the design features of APR1400, an evolutionary pressurized water reactor developed by the Korean industry. ATLAS was used to perform a set of integral effect tests on the reflood period of a large-break loss-of-coolant accident (LBLOCA) after intensive performance tests had been conducted to verify ATLAS’s operational performance and controllability for major thermal-hydraulic components. The present LB-CL-09 test is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period that can be used to provide reliable data to help validate the LBLOCA analysis methodology for APR1400. The main objective of the present test is to identify the major thermal-hydraulic characteristics such as the direct emergency core coolant (ECC) bypass, downcomer boiling, and core cooling behavior during the reflood phase of an LBLOCA for APR1400 under conditions where the downcomer region interacts with the reactor core region and the heat could be transferred through the steam generator. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results. The decay heat and the ECC flow rate from the safety injection tank were simulated from the start of the reflood period. The ECC flow rate from the safety injection pump was 0.32 kg/s. The system pressure was fixed at ~0.1 MPa, and the initial outer-wall temperature was determined to be 205°C. The experimental results showed the typical thermal-hydraulic trends expected to occur during the reflood phase of the LBLOCA scenario.

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

For AHWR (Advanced Heavy Water Reactor), a pressure tube type Boiling Water Reactor (BWR) with parallel inter-connected loops, the Steam Drum (SD) level control is closely related to Main Heat Transport (MHT) coolant inventory and sustained heat removal through natural circulation, hence overall safety of the power plant. The MHT configuration with multiple (four) interconnected loops influences the SD level control in a manner which has not been previously addressed. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Event (PIEs) for Loss of Coolant Accident (LOCA), which postulates a double ended break in the four partitioned Emergency Core Cooling System (ECCS) header. A conventional individual three-element SD level controller can not account for the highly coupled and interacting behaviors, of the four SD levels. An innovative three-element SD level control scheme is proposed to overcome this situation. The response obtained for a variety of unsymmetrical disturbances shows that the SD levels do not diverge and quickly settle to the various new set points assigned. The proposed scheme also leads to enhanced safety margins for most of the PIEs considered with a little influence on the 100% full power steady-state design conditions.

RISK INFORMED DECISION MAKING USING MULTIOBJECTIVE OPTIMIZATION TECHNIQUE

Nuclear Power Plant operations are guided by Limiting conditions of operations (LCO's) laid out in the document referred to as the Technical Specifications (TS). This Technical Specification is a legitimate framework approved by the Regulatory Bodies for the Safe operations of the Nuclear Power plants. In the past, the regulatory bodies used a deterministic approach as the basis for making decisions on safety issues and organizing the activities that they carry out. This was done by applying high level criteria such as the need to provide defence in depth and adequate safety margins. However with the availability of detailed Plant Specific Level-1 Probabilistic Safety Assessment (PSA), these limiting conditions need to be reviewed/revised based on the analysis results. This review of the LCO's is not a trivial exercise if the entire solution space of the variables defining the variables has to be investigated. This paper reviews the case for revision of Surveillance test frequencies of the Emergency Core Cooling System injection valves, using the multiobjective optimization technique.

AN AXIOMATIC DESIGN APPROACH OF NANOFLUID-ENGINEERED NUCLEAR SAFETY FEATURES FOR GENERATION III+ REACTORS

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.

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.

CFD Modeling and Experiments of Insulation Debris Transport Phenomena in Water Flow

The investigation of insulation debris generation, transport, and sedimentation becomes more important with regard to reactor safety research for pressurized water reactors and boiling water reactors when considering the long-term behavior of emergency core coolant systems during all types of loss-of-coolant accidents (LOCAs). The insulation debris released near the break during a LOCA incident consists of a mixture of disparate particle populations that varies with size, shape, consistency, and other properties. Some fractions of the released insulation debris can be transported into the reactor sump, where it may perturb/impinge on the emergency core cooling systems.Open questions of generic interest are, for example, the particle load on strainers and corresponding pressure drop, the sedimentation of the insulation debris in a water pool, and its possible resuspension and transport in the sump water flow. A joint research project on such questions is being performed in cooperation with the University of Applied Sciences Zittau/Görlitz. The project deals with the experimental investigation and the development of computational fluid dynamics (CFD) models for the description of particle transport phenomena in coolant flow. While the experiments are performed at the University of Applied Sciences Zittau/Görlitz, the theoretical work is concentrated at Forschungszentrum Dresden-Rossendorf.In the current paper the basic concepts for CFD modeling are described and feasibility studies including the conceptual design of the experiments are presented.

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.

Two‐Phase Flow Simulations for PTS Investigation by Means of Neptune_CFD Code

Two‐dimensional axisymmetric simulations of pressurized thermal shock (PTS) phenomena through Neptune_CFD module are presented aiming at two‐phase models validation against experimental data. Because of PTS complexity, only some thermal‐hydraulic aspects were considered. Two different flow configurations were studied, occurring when emergency core cooling (ECC) water is injected in an uncovered cold leg of a pressurized water reactor (PWR)—a plunging water jet entering a free surface, and a stratified steam‐water flow. Some standard and new implemented models were tested: modified turbulent k‐ε models with turbulence production induced by interfacial friction, models for the drag coefficient, and interfacial heat transfer models. Quite good agreement with experimental data was achieved with best performing models for both test cases, even if a further improvement in phase change modelling would be suitable for nuclear technology applications.

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.

Air–water multidimensional impingement flow on vertical flat wall

The emergency core cooling (ECC) water is supplied from the direct vessel injection (DVI) system in the Advanced Power Reactor 1400 MWe (APR1400) during a postulated large-break loss-of-coolant accident (LBLOCA). The velocity of ECC water exceeds 10 m/s in the early high pressure phase of LBLOCA and then is decreased to 2–3 m/s in the late phase of reflood. During the injection the flow behavior exhibits a complex mode involving impingement, bypass, entrainment, sweepout and condensation in the reactor downcomer. There is currently no model to accurately simulate the local and complicated flow behavior in the APR1400 downcomer during a LBLOCA. This study is aimed at developing models for the water film flow and deformation, both of which are expected to sizably affect the other multidimensional flow characteristics in the downcomer. Experimental studies are conducted to benchmark the predictive model by furnishing the boundary conditions for the analysis resorting to the Accelerated Liquid Phase Hydrodynamics Apparatus (ALPHA) and the Kinetic Aerodynamic Physics Parallelepiped Apparatus (KAPPA). The Poisson equation and potential theory are applied to formulate the behavior of the water film and air flow. In both the experimental and numerical studies, the temperature-dependent thermodynamic properties and the reactor vessel curvature are neglected to render the problem at hand tractable. The model is found to reasonably describe the downward film flow behavior. The water film is developed in proportion to the initial injection velocity of the ECC water. The downward velocity of water film is increased with the heights of injection. Regarding the film deformation the calculated results tend to deviate from the experimental data as the injected air velocity is increased. The disagreement is attributed to limitations inherent in the two-dimensional treatment and point source approach.

Implementation of a strainer model for calculating the pressure drop across beds of compressible, fibrous materials

Mineral wool insulation debris, which is generated during a loss-of-coolant-accident (LOCA), has the potential to undermine the long-term recirculation capability of the emergency core coolant system (ECCS) in a nuclear power plant. Most importantly, ECCS pumps are faced with an increasing pressure drop while insulation debris accumulates at the pump suction strainers. The presented study aims at modelling the pressure drop of flows across growing cakes of compressible, fibrous materials and at the implementation of the model into a general-purpose three-dimensional (3D) computational fluid dynamics (CFD) code. Computed pressure drops are compared with experimentally found values. The ability of the CFD implementation to simulate 3D flows with a non-uniformly distributed particle phase is exemplified using a step-like channel geometry with a horizontally embedded strainer plate.

Numerical and experimental investigations for insulation particle transport phenomena in water flow

The investigation of insulation debris generation, transport and sedimentation becomes more important with regard to reactor safety research for pressurized and boiling water reactors, when considering the long-term behaviour of emergency core coolant systems during all types of loss of coolant accidents (LOCA). The insulation debris released near the break during a LOCA incident consists of a mixture of a disparate particle population that varies with size, shape, consistency and other properties. Some fractions of the released insulation debris can be transported into the reactor sump, where it may perturb or impinge on the emergency core cooling systems. Open questions of generic interest are for example the particle load on strainers and corresponding pressure-drop, the sedimentation of the insulation debris in a water pool, its possible re-suspension and transport in the sump water flow. A joint research project on such questions is being performed in cooperation with the University of Applied Science Zittau/Görlitz and the Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation and the development of computational fluid dynamic (CFD) models for the description of particle transport phenomena in coolant flow. While the experiments are performed at the University Zittau/Görlitz, the theoretical work is concentrated at Forschungszentrum Dresden-Rossendorf. In the present paper, the basic concepts for computational fluid dynamic (CFD) modelling are described and experimental results are presented. Further experiments are designed and feasibility studies were performed.

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.

Development of passive flow controlling safety injection tank for APR1400

A vortex valve, called fluidic device, is to be installed inside a Safety Injection Tank (SIT) of Advanced Power Reactor 1400 MWe (APR1400) that passively controls an Emergency Core Cooling (ECC) water discharge flow rate without any moving part or any action of the plant operator. The fluidic device was designed, and its performance was evaluated by a series of repetitive experiments using VAlve Performance Evaluation Rig (VAPER), a prototypical full-scale test facility. The passive flow controlling SIT satisfied the major performance requirements of the APR1400 plant design, in view of peak discharge flow rate, pressure loss coefficient, and duration time of the ECC water discharge.

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.