Main Coolant Pump Research Articles

High-Temperature and High-Pressure Tribological Properties of Siliconized Graphite for Water-Lubricated Thrust Bearing Application in Main Coolant Pump

The effect of the microstructure of siliconized graphite on tribological properties is investigated by using a high-temperature and high-pressure water-lubricated tribometer on a self-mated ring-on-ring configuration under an applied load of 500–1500 N with a spindle speed of 100–5000 rpm in both 90 °C (5 MPa) and 25 °C (1 MPa) water environments, respectively. The Stribeck curves measurement and continuous wear tests are performed and analyzed in both water environments. The wear behaviors of the graphite, SiC, and free-silicon phases in siliconized graphite are demonstrated to explore the wear mechanism. The larger wear depths of a low-worn surface roughness on the three phases contribute to the boundary lubrication. The shallower wear depths are observed on the SiC and Si phases under the mixed lubrication, corresponding to partial contact wear of surface asperities. The wavy surface of the SiC phase and uniform flow-oriented striae of the Si phase are attributed to hydrodynamic lubrication, caused by full water film scouring the worn surface. Finally, an integrated evaluation method of G duty parameters is successfully used to identify the lubrication regimes of siliconized graphite from the boundary, mixed, to hydrodynamic lubrications for a water-lubricated thrust bearing application in the main coolant pump of a nuclear power plant.

Development and benchmarking of SIMULATE5-K-VVER

SIMULATE5-K-VVER (S5K-VVER) is the newest addition to the CMS5-VVER modeling suite developed by Studsvik Scandpower, Inc. (SSP). In S5K-VVER, the coupled time-dependent multigroup neutron diffusion equation, delayed neutron precursor concentrations, and thermal hydraulic models are solved in triangular nodal geometry to model transients in Water-Water Energetic Reactors (VVERs).The neutron kinetics equations in S5K-VVER are solved in a two-step process: temporal integration and spatial integration. In S5K-VVER there are two options for temporal integration: the Frequency Transform Method (FTM) and a Backward-Euler Method (BEM).S5K-VVER is then used to solve a series of benchmark problems including the AER-DYN-001 and AER-DYN-002 numerical transient benchmarks. Finally, the Kalinin-3 benchmark is modeled which describes the shutting-off of a single Main Coolant Pump (MCP) in the Kalinin-3 VVER-1000/320 nuclear power plant. Compared to plant data, the S5K-VVER solution agrees within 1.5% of core total power during this transient.

Study on the hydraulic performance of the axial flow main coolant pump for lead-cooled fast reactor

Lead-cooled fast reactor (LFR) has been listed as the most promising fourth-generation reactor type for commercial applications. LFR uses liquid heavy metal, lead or lead–bismuth eutectic alloy (LBE) as the coolant. It has completely different physical parameters such as the density and viscosity from the traditional nuclear reactor coolant, water. Different physical parameters will have anoticeableeffect on the hydraulic performance of the main coolant pump. In this study, based on SST k-w turbulence model, an axial flow main coolant pump in the LFR at multiple working conditions was simulated, and the differences of its hydraulic performance under LBE and water and the reasons were analyzed in detail. Both the head and efficiency of the axial flow pump model under LBE are higher obviously than that under water, and the theoretical head is basically the same under the two mediums. The increase of the head and efficiency under LBE mainly results from the decrease of profile losses in the impeller and guide vane, both the friction loss and losses due to vortices in the boundary layer and wake formation at the outlet of the blade are obviously smaller than that under water.

Validation of the Ants-TRACE code system with VVER-1000 coolant transient benchmarks

The objective of this paper is to start the validation of the coupled code system of the nodal neutronics solver Ants and the system code TRACE. The process consists of two exercises of well-known VVER-1000 coolant transient benchmarks: a core-vessel simulation of a main steam line break transient of the V1000CT-2 benchmark and a full plant simulation of a main coolant pump trip of the Kalinin-3 benchmark. The Ants-TRACE results are compared against other code solutions in the V1000CT-2 benchmark and measurement data in the Kalinin-3 benchmark. The results show good agreement in both benchmarks, with the deviations ranging mostly within the measurement error or the range of deviations of other published solutions. The results confirm the correct implementation of the coupling and successful modeling of relevant phenomena in coolant transient events in a VVER-1000 reactor.

ATHENA Main Heat Exchanger Conceptual Design and Thermal-Hydraulic Assessment With RELAP5 Code

Abstract Among the envisaged experimental infrastructures supporting ALFRED reactor development, Fostering Alfred Construction consortium identified ATHENA as one of the facilities to address the pool thermal-hydraulic challenges and demonstrate the feasibility of the revised ALFRED configuration, along with the thermal-hydraulic performances of its main components. ATHENA is a large pool-type lead-cooled multipurpose experimental facility featuring a large size vessel (3.2 m diameter, 10 m in height), conceived to host almost 800 tons of lead to test ALFRED relevant scaled components. The test section to be installed in the main vessel includes an electrically heated core simulator (CS), made of 7 FAs, which delivers to the primary coolant a nominal thermal power of 2210 kW, a main coolant pump for lead circulation, and a countercurrent shell and tubes main heat exchanger, which tube bundle is fed by pressurized water by a dedicated secondary circuit. A preliminary analytical calculation has been performed to identify the most suitable configuration for the ATHENA heat exchanger, as well as to achieve a preliminary sizing of the component. The final layout foresees a bundle of 91 bayonet tubes, arranged in a cylindrical geometry, having an active length of 3000 mm and adopting an external double wall tube with a pressurized helium gap. A numerical model of the heat exchanger has been developed using the thermal-hydraulic system code RELAP5/Mod3.3, and a numerical sensitivity analysis on the geometrical and operating parameters has been carried out to verify the performances of the component.

Thermal-Hydraulic Characterization and Numerical Modeling with RELAP5 Code of ATHENA Secondary Loop

Among the envisaged experimental infrastructures supporting Advanced Lead Fast Reactor European Demonstrator (ALFRED) reactor development, the FALCON [Fostering ALfred CONstruction] consortium identified the Advanced Thermo-Hydraulics Experiment for Nuclear Application (ATHENA) as one of the facilities to address the pool of thermal-hydraulic challenges and demonstrate the feasibility of the revised ALFRED configuration, along with the thermal-hydraulic performances of its main components. ATHENA is a large pool-type lead-cooled multipurpose experimental facility featuring a large-sized vessel (3.2-m diameter and 10-m height), conceived to host almost 800 tons of lead to test ALFRED-relevant scaled components. The test section to be installed in the main vessel includes an electrically heated core simulator, made of seven fuel assemblies, which delivers to the primary coolant a nominal thermal power of 2210 kW; a main coolant pump for lead circulation; and a countercurrent shell-and-tube main heat exchanger, of which the tube bundle is fed by pressurized water by a dedicated secondary circuit. This work presents the numerical model of ATHENA along with thermal-hydraulic characterization of the facility using the system code RELAP5/Mod3.3, properly modified to include the thermophysical properties of heavy liquid metals. After the characterization of the steady state representative of the Stage 3 foreseen for the ALFRED staged approach, results of a numerical sensitivity analysis aimed at defining the most suitable procedure for the shutdown transient of the facility are presented.

Thermal-Hydraulic Transient Analyses for Primary Coolant System of HFRR Using RELAP5

The High Flux Research Reactor (HFRR) is a 40 MW pool-type nuclear research reactor developed for neutron beam tube research, material and fuel irradiation studies, and advanced research using a cold neutron source. As a part of safety analyses, thermal-hydraulic transient analyses were carried out using the computer coder RELAP5 for the primary coolant system of the HFRR. This technical note presents various analyses that were carried out such as loss of normal electric power supply, loss of heat sink, main coolant pump seizure, inadvertent opening of a natural circulation valve, and loss-of-coolant accident considering breaks in the pump suction line and pump discharge line. It is demonstrated that fuel cooling is ensured and safety parameters are well within acceptable limits.

Study on the Hydraulic Performance of the Axial Flow Main Coolant Pump for Lead-Cooled Fast Reactor

Study on the Hydraulic Performance of the Axial Flow Main Coolant Pump for Lead-Cooled Fast Reactor

Transient analysis of a locked rotor/shaft seizure accident involving the EU-DEMO WCLL Breeding Blanket primary cooling circuits

The EU-DEMO Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) main subsystems to be cooled are the Breeder Zone (BZ) and the First Wall (FW). Each subsystem will be equipped with an independent Primary Heat Transfer System (PHTS). Within the framework of the EUROfusion Work Package Breeding Blanket research program, several accidents belonging to the category of “Decrease in Coolant System Flow Rate” were studied. The activity was aimed at evaluating the blanket and primary cooling systems thermal-hydraulic performances during such transient conditions. A complete model including the BB and related PHTS circuits has been developed at Sapienza University of Rome. A modified version of RELAP5/Mod3.3 system code has been used to perform the calculations. The simulation results showed that a locked rotor/shaft seizure of a BZ or a FW main coolant pump is the most challenging scenario. BZ and FW system behavior has been analyzed following this initiating event with the goal of the design improvement and to individuate the need for preventive measures. The influence of loss of off-site power on the accident evolution has also been investigated. Moreover, management strategies have been proposed for different reactor components. Calculations demonstrate that the current blanket and PHTS design is appropriate to cope with these kinds of accident scenarios.

Loss of flow accident analysis in Atucha II nuclear power plant using RELAP5 model

The loss of flow accident (LOFA) in nuclear power plants (NPP) is considered a design base initiating event (DBIE) and it is broadly analyzed in safety and licensing evaluation. Therefore, this is commonly chosen to assess computational models. The current paper addresses the simulation of LOFA under single and total main coolant pump (PMP) trips during full power in Atucha-II Pressurized Heavy Water Reactor (PHWR). RELAP5-mod3.3 NPP was used for solving the thermal-hydraulics, neutron kinetics, and control and safety logic in steady-state and LOFA events. Simulations were assessed against real plant data recorded during a single-PMP trip obtaining very good agreement. After that, the total-PMP trip was simulated, showing that the plant also reached a safe steady condition, guaranteeing the removal of decay heat by natural flow circulation. The very good agreement in terms of the main thermal-hydraulics parameters demonstrated the capability of the computational model to reproduce the overall NPP behavior.

Transient trend prediction of safety parameters for small modular reactor considering equipment degradation

Small modular reactor (SMR) has strong coupling characteristics and complex nonlinearities. Predicting the trend of SMR safety parameters and monitoring whether the parameter values exceed operational limits and conditions (OLC) before performing a transient helps the operator to judge the safety of the transient, especially in the case of equipment degradation. In this paper, a transient trend prediction method of SMR safety parameters based on random forest model was proposed, which considered the impact of equipment degradation on safety parameters. The transient data used for model training included different load-following conditions considering various degraded degrees of rotating speed drop in the main coolant pump (MCP) or effective heat transfer area reduction in the once-through steam generator (OTSG). To evaluate the model, the new untrained transient data was used as test data, the results showed that safety parameters were more accurately predicted by random forest, compared with the backpropagation neural network.

Модель реактивнісної аварії на РБМК-1000

The reactor model was used to study the accident that destroyed the RBMK-1000 reactor at Unit 4 of the Chornobyl nuclear power plant on 26 April 1996. The model of reactivity accident of the RBMK-1000 reactor is based on equations of nuclear reactor kinetics, taking into account feedback in reactor reactivity. Reactivity changes as a result of both external influences – the movement of regulatory organs, changes in the reactor inlet coolant temperature, – and as a result of feedback by core parameters – changes in fuel temperature, coolant density, and 135Хе concentration. The model takes into account steam generation in the reactor core, which corresponds to the real physics of processes at the RBMK reactor, and allows obtaining simulation results that best match the recorded data and the consequences of the accident process. The study of reactivity accident on RBMK-1000 reactor is carried out for different combinations of values of control rods efficiency; reactivity coefficients by fuel temperature and coolant density; changes in the reactor inlet coolant temperature; the emergency protection time, as well as the reactor power level before closing the turbine generator stop valve. Different reactivity accident scenarios at RBMK-1000 reactor allow us to determine the most unfavorable combinations of external influences on the course of reactivity accident, namely: start time of main coolant pump rundown, time of activation of emergency protection, power level before the closing of turbine generator stop valves. In most reactivity accident scenarios, first of all, the critical values of fuel enthalpy are reached, at which the process of fuel destruction in the fuel element, destruction of the fuel assembly, and assembly channel start. Important results of studies are 1 – determination of the fact that time of activation of emergency protection after the closing of stop valves of turbine generator significantly affects the value of the maximum neutron power that is achieved during a reactivity accident; 2 – determination of the effect of reactor power before the closing of turbine generator stop valves on the course of the accident; 3 – it is not necessary to achieve supercritical on instantaneous neutrons, supercritical on delayed neutrons is enough to start fuel destruction.

Shutdown idling performance of the nuclear main coolant pump under station blackout accident: An optimization study

Lead-cooled fast reactor (LFR) is an important reactor type of the Generation IV nuclear energy, and the main coolant pump (MCP) is the only and most critical power equipment in primary circuit of LFR. The shutdown idling performance of the MCP, is one of the most important safety indicators in the event of a station blackout accident. When the motor loses its power supply, the MCP needs to run under the inertia of the flywheel to ensure the coolant flow for a short time. This study mainly focuses on the idling mathematical model of the MCP under power failure accidents. In this study, taking the shutdown idling performance of the MCP as the optimization goal, the parametric optimization analysis of the impeller and diffuser structure of the MCP was carried out. Specifically, based on the platform ISIGHT, the software of CFturbo, TurboGrid, CFX, Matlab and FLOWMASTER are systematically integrated, the automated 3D modeling, automatic meshing and automatic CFD calculation of different hydraulic model of the MCP are realized. And the neural network mathematical model between the geometric parameters of MCP and the idling performance indicators is established. The study shows that there is a strong linear relationship between the idling performance of the MCP and its head and shaft power, and meridian parameters of Δβ 2, Z 1, Φ 1h and the radial surface parameters of Δβ 1 and Δβ 2, Z 1 and Φ 1h have the greatest impact on the idling performance of MCP.

Simulation of Partial and Complete Loss of Flow Accidents for GPWR using ATHLET Code

In Nuclear power plants, simulation of the thermal hydraulic systems at loss of flow, accidents is commonly used to perform safety and licensing analyses through regulatory and safety authorities since they are considered as design base accidents. In this paper, a simulation of a thermal-hydraulic model of Germany Pressurized Water Reactor (GPWR) is developed, during loss of flow accident (LOFA) scenarios, such as: trip of one, two, three and four of the Main Coolant Pumps, at operation of 100% reactor power. The developed thermal hydraulic model is formed using ATHLET3.1A code with a point reactor kinetic to predict the behavior of the reactor through the thermal hydraulic major parameters such as power, mass flow rate, primary and secondary pressures, max clad surface and fuel temperatures. All reactivity feedbacks are taken into consideration in the developed simulation and the scram signal of the protection system depends on the mass flow rate. The results show that the investigated ATHLET model can precisely simulate, the steady state and the transient behavior of GPWR during many LOFA accidents. Moreover, the results show that there is good agreement between ATHLET model and RELAP5 code results are nearly in agreement with in previous literatures

Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category

The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes.

Comparative analysis of internal flow characteristics of LBE-cooled fast reactor main coolant pump with different structures under reverse rotation accident conditions

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

Analysis of the thermal-hydraulic behavior of the EU-DEMO WCLL breeding blanket cooling systems during a loss of flow accident

Abstract The water-cooled EU-DEMO Breeding Blanket (BB) is cooled by two independent circuits, the Breeder Zone (BZ) and the First Wall (FW) Primary Heat Transfer Systems (PHTS). The configuration under study foresees the presence of an Intermediate Heat Transfer System and an Energy Storage System to operate the turbine during both the pulse (2 h) and the dwell time (10 min) at almost constant load, despite the plasma power pulsation. Within the framework of the EUROfusion WPBOP research activity, a RELAP5/Mod3.3 model was developed to investigate the thermal-hydraulic behavior of the primary cooling systems during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”. The nodalization includes the BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat EXchangers. The model was initially used to simulate the nominal conditions with the pulse and dwell phases. Then, starting from the pulse, a Loss of Flow Accident (LOFA) was selected to preliminary evaluate the PHTS behavior with the aim of the design improvement. LOFA analyses were performed considering the complete loss of both the FW and BZ PHTS main coolant pumps (MCPs). A sensitivity was carried out to assess the impact of the MCPs flywheel on the main PHTS parameters. Transient results highlighted the appropriateness of the current design with no need for further mitigation actions.

Dynamics of impulse seals with tubular feeders

The paper presents a technique for calculating forced axial oscillations and stability of impulse seals with tubular feeders. Unlike the existing methods of calculation, wherein, for the sake of simplicity, the influence of the feeder conductance is neglected, this article describes a conductance effect on the dynamic characteristics of the impulse face seal. The proposed calculation technique makes it possible to determine amplitudes of forced oscillations of the axially movable bushing and stability limit for the geometric parameters chosen in the static calculation. As an example, the authors described a calculation of the impulse face seal of the rotor of the main Reactor Coolant Pump (RCP 20000-100). The amplitude and phase-frequency characteristics for different values of the nominal gap are constructed.

Research on structure selection and design of LBE-cooled fast reactor main coolant pump

Fast neutron reactor has become the preferred reactor type for advanced nuclear energy systems in the world, representing the development direction of the Generation IV nuclear energy system. The lead-cooled fast reactor (LFR) uses pure Pb/LBE as coolant, the natural characteristics of lead alloy are combined with the structural design of LFR, which constitutes the inherent safety characteristics of LFR and the system. In this paper, the main coolant pump in the Generation IV of LFR is taken as the research object. According to the design requirements of the main coolant pump in LFR fast reactor, the hydraulic characteristics of the main coolant pump with double-outlet volute type and space guide vane type under normal operation and accident operation are compared and analyzed. It is found that, under normal pump conditions the hydraulic performance of double-outlet volute type main coolant pump is a slightly better, but the space guide vane type has obvious advantages in the ability to deal with accident conditions and the better reliability in its long-term operation. Finally, the space guide vane structure is determined as the design program of the main coolant pump of CLEAR-I fast reactor.

SKETCH-N/ATHLET coupled calculations using the boundary conditions plugin

The paper describes recent developments in the multi-physics coupling scheme between the SKETCH-N nodal neutronics code and the best-estimate thermohydraulic code ATHLET v3.2. The boundary conditions plugin was implemented. The verification and validation were performed using the transient of the Kalinin-3 international Benchmark. The simulation results using the boundary conditions plugin show good agreement with experimental data and calculations performed by using SKETCH-N/ATHLET direct calculations. The calculations of the transient “Switch off of one MCP (Main Coolant Pump) at nominal power” is performed applying a simple core thermohydraulic model without taking into account inter-channel mass transfer.