Main Circulation Pumps Research Articles

Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility

ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility under construction at the Mioveni site in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to characterize large to full-scale ALFRED components, to conduct integral tests, and to investigate the main thermal–hydraulic phenomena inherent in pool-type systems. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. Similar to ALFRED’s design, ATHENA minimizes thermal stratification within the main vessel even under natural circulation conditions, through an internal structure referred to as “barrel”. This structure directs the fluid flow towards the main vessel, preventing fluid stagnation near the vessel itself. The paper initially provides a steady-state thermal–hydraulic characterization of the facility, including details of the numerical model developed using the RELAP5/Mod3.3 thermal–hydraulic code. Then, focus is given to the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient. In this scenario, the Main Circulation Pump (MCP) is assumed to remain operational while the Core Simulator (CS) is deactivated once the lead temperature at the Main Heat Exchanger (MHX) outlet reaches a predefined threshold. A sensitivity analysis is conducted with set points of 430 °C, 450 °C, 470 °C, and 490 °C, assessing the system’s response following MHX isolation from the secondary loop. The study evaluates the impact of different CS deactivation set points on reactor SCRAM delay (reducing CS power to a level representative of decay heat) as well as on system maximum and minimum temperatures.

Development and validation of transient analysis module in nodal diffusion code RAST-V with Kalinin-3 coolant transient benchmark

This study introduces a transient analysis module developed for RAST-V and validates it using the Kalinin-3 benchmark problem. For the benchmark analysis, RAST-V standalone and STREAM/RAST-V calculations were performed. STREAM supplies the few-group constants and RAST-V conducts a 3D core simulation utilizing few-group cross-sectional data. To improve accuracy, the main solver was developed based on the advanced semi-analytic nodal method. To evaluate the computational capability of the transient analysis module in RAST-V, Kalinin-3 benchmark is employed. Kalinin-3 represents a coolant transient benchmark that offers experimental data during the deactivation of the Main Circulation Pumps. Consequently, the transient calculations reflected the changes in the reactor flow rate. Benchmark comprising steady-state and transient calculations. During the steady state, the STREAM/RAST-V combination demonstrated a 30 ppm root mean square difference from 0 to 128.50 EFPD. For the transient calculations, STREAM/RAST-V showed power differences within ±7 % over a range of 0–300 s. Axial offset differences were within ±3 %, and the RMS difference in radial power ranged within 2.596 % at both 0 and 300 s. Overall, this study effectively demonstrated the newly developed transient solver in RAST-V and validated it using the Kalinin-3 benchmark problem.

On the issue of carrying out mooring tests of a ship power plant in shallow water conditions

In the process of mooring trials, as well as when ships are sailing in shallow water, difficulties often arise in ensuring normal operating conditions for the propulsion complex, the performance of the power plant and the ship as a whole. The article outlines the problems during mooring tests in shallow water. Mooring tests of a large-tonnage tanker in shallow water conditions are considered, when the operation of the main circulation pump of the cooling system is disrupted due to contamination of the flow path of the main condenser with soil particles. The economic inexpediency of testing a tanker at sea when it is anchored or barreled in an area that is safe for navigation and has sufficient water depth is substantiated. The method of taking cooling water from above with the help of a caisson is considered, which eliminated unproductive costs. A method of mooring a ship is given, in which the water jet from the propeller does not harm the berthing facilities and does not erode the soil of the water area; the means of simulating the nominal parameters of the actuators are described, which makes it possible to carry out mooring tests of mechanisms, systems and devices without bringing the main power plant to the mode of increased loads, and, as a result, to reduce the testing time.

Noise Reduction Method for the Vibration Signal of Reactor CRDM Based on CEEMDAACN-SK

The reactor control rod drive mechanism (CRDM) controls the startup, shutdown and power of the reactor; it is one of the key pieces of equipment to ensure the normal operation of the reactor. CRDM is complex, mainly composed of stator, rotor, bearing, roller, etc. The characteristic analysis of the vibration monitoring signal is one of the important methods of the CRDM state evaluation. In view of the characteristics of large noise interference and the difficulty in analyzing the vibration monitoring signal of CRDM, this paper proposes a noise reduction method for the vibration signal of CRDM based on complete ensemble empirical mode decomposition with adaptive amplitude correction noise and spectral kurtosis (CEEMDAACN-SK), which can deeply reduce the vibration signal of CRDM. Firstly, the proposed CEEMDAACN algorithm is used to decompose the vibration signal of CRDM to obtain multiple intrinsic mode functions (IMF). Then, the spectral kurtosis of each IMF component is analyzed to obtain the spectral kurtosis map of each IMF component, which is compared with the spectral kurtosis map of the original signal. Finally, the denoising reconstruction of the signal is carried out to obtain the final denoising signal. Through experimental analysis, the performance of the proposed CEEMDAACN-SK denoising algorithm is better than the complete ensemble empirical mode decomposition with adaptive noise and spectral kurtosis (CEEMDAN-SK) algorithm in terms of results. The method proposed in this paper can be applied not only to the vibration signal noise reduction of CRDM, but also to other equipment and fields, such as nuclear power main circulation pump and the chemical industry.

Digital acoustic model of an NPP pressurizer with WWER

It is experimentally proven that the main cause of vibration excitation of equipment and internals of the main circulation circuit (MCC) are acoustic standing waves (ASW) and main circulation pumps (MCP). Usage of an interdisciplinary approach made it possible to create a digital acoustic model of pressurizer system with pipelines attached to it and prove that it is a self-oscillating system capable of generating acoustic standing waves (ASW) similar to simultaneous operation of several Helmholtz resonators. Each Helmholtz resonator in pressurizer is able to suppress a certain frequency of ASW generated by the reactor, which depends on coolant temperature. Determination of natural frequencies of coolant pressure oscillations (NFCPO) in pressurizer using Thomson formula is based on method of electroacoustic analogy. Acoustic compliance of pressurizer and acoustic mass of various combinations of pipelines connected to it make it possible to calculate NFCPO, in the system under consideration, according to the formula presented in the form of acoustic analogues of inductance and capacitance. Acoustic parameters of a system of Helmholtz resonators formed by a pressurizer and a connecting pipeline with hot leg of the third loop in nominal operating mode of NPP with WWER-1000 are calculated. A technology has been developed and verified, advantage of which is usage of a pressurizer for damping ASW, frequencies of which fall within vibration bandwidth of primary circuit structural elements. Calculation and experimental proof of previously unknown ability of pressurizer to damp ASW generated by the reactor in a wide frequency range is given, and possibility of preventing resonances with vibrations of equipment and NPP structures is substantiated.

Special features of assessing the strength of nuclear reactor equipment under dynamic loads

The paper presents tasks and approaches to their solutions in assessing vibration strength and dynamic safety of marine reactor equipment (for example: pumps, control driven rod mechanisms). The topicality of this work is determined by the experience of operating nuclear power plants. Development of digital computation models and computational analysis allowed us to exclude experimental validation of control driven rod mechanisms seismic resistance. A digital computational model was developed for validating the dynamic safety and life time of the seal pot of the reactor facility main circulation pump. The model included the main circulation pump and the framework of the reactor facility. Computational and experimental analysis of the vibration state of the control and protection system drives was also carried out.

A Novel Ensemble Fault Diagnosis Model for Main Circulation Pumps of Converter Valves in VSC-HVDC Transmission Systems.

The intelligent fault diagnosis of main circulation pumps is crucial for ensuring their safe and stable operation. However, limited research has been conducted on this topic, and applying existing fault diagnosis methods designed for other equipment may not yield optimal results when directly used for main circulation pump fault diagnosis. To address this issue, we propose a novel ensemble fault diagnosis model for the main circulation pumps of converter valves in voltage source converter-based high voltage direct current transmission (VSG-HVDC) systems. The proposed model employs a set of base learners already able to achieve satisfying fault diagnosis performance and a weighting model based on deep reinforcement learning that synthesizes the outputs of these base learners and assigns different weights to obtain the final fault diagnosis results. The experimental results demonstrate that the proposed model outperforms alternative approaches, achieving an accuracy of 95.00% and an F1 score of 90.48%. Compared to the widely used long and short-term memory artificial neural network (LSTM), the proposed model exhibits improvements of 4.06% in accuracy and 7.85% in F1 score. Furthermore, it surpasses the latest existing ensemble model based on the improved sparrow algorithm, with enhancements of 1.56% in accuracy and 2.91% in F1 score. This work presents a data-driven tool with high accuracy for the fault diagnosis of main circulation pumps, which plays a critical role in maintaining the operational stability of VSG-HVDC systems and satisfying the unmanned requirements of offshore flexible platform cooling systems.

Numerical Simulation of the Welding Process during the Main Circulation Pump Impeller Repair

Paper presents a numerical simulation of the main circulation pump impeller damaged parts repair/replacement welding process, for the prediction of welding-induced residual stresses. 3D numerical model creation of the impeller flange and central hub welded joints. Thermal and stress-strain analysis of assessed parts using the finite element method. Calculation of equivalent residual stresses in the area of examined welded joints. Applicability of results in further impeller integrity analyzes. The results confirmed the suitability of the chosen welding procedure and technology.

Influence of Surface Machining on Development of Stress Corrosion Cracking of Distribution Wheels of Main Circulation Pump in Nuclear Power Plants

The present work deals with the causes of the damage to the distribution wheel of the main circulation pump of the nuclear power plant. During the solution of the problem, a significant influence of the final machining on the formation of cracks was found. Coarse surface machining can cause strain-induced martensite on the distribution wheel surface. Small pitting, which was observed on the surface of the distribution wheel, can be caused by the presence of martensite with less corrosion resistance. Thermal stress and residual stress after coarse final surface machining caused the growth of the cracks which initiated from the pitting.

Health State Identification Method of Nuclear Power Main Circulating Pump Based on EEMD and OQGA-SVM

Main circulation pump is the only high-speed rotating equipment in primary loop of nuclear power plant. Its function is to ensure the normal operation of primary loop system by controlling the circulating flow of reactor coolant. In order to ensure long-term healthy operation of nuclear power main circulating pump, a method for identifying the health states of nuclear power main circulating pump based on ensemble empirical mode decomposition (EEMD) and support vector machine optimized by optimized quantum genetic algorithm (OQGA-SVM) is proposed. Vibration signal of main circulating pump is decomposed by EEMD. Vibration signal characteristics of nuclear power main circulating pump in healthy state and different fault states are analyzed and target characteristic indexes are put forward. Then, health state identification model of main circulation pump of OQGA-SVM is established, and target characteristic indexes are used as input parameter of the model. Finally, combined with experimental data, the model analysis and validation show that the health state identification method of nuclear power main circulating pump based on EEMD-OQGA-SVM can accurately and effectively identify the states of main circulation pump, has a higher identification accuracy than EEMD-SVM method and is more efficient and accurate than EEMD-QGA-SVM method.

Experience of using loose parts monitoring systems at Novovoronezh NPP

In VVER reactor plants, it is impossible to completely exclude the appearance of loose, loosely fixed and foreign objects in the main circulation circuit. Operational experience shows that early detection and estimation of the parameters of such incidents can provide the time required to eliminate or minimize damage to the main equipment of the reactor plant. For this reason, most modern power units with pressurized water reactors (PWR, VVER) are equipped with a loose parts monitoring system (LPMS). At the units under construction, these systems are laid down as standard ones; the power units put into commercial operation in the Soviet period were also equipped with them. The requirements for them are established by international standards. Ongoing research work in this area is aimed at determining the root cause of the acoustic anomaly and the localization of its epicenter. Also, no less significant are the works aimed at determining the mass of a loose object (LO). The most precise definition of this parameter will make it possible to have an idea of the nature of the LO before its withdrawal from the primary circuit and to conclude about whether this object is accidentally found or it is a detached part of the steam generators, main circulation pumps, internal devices or shut-off and control valves.

Two-stage modelling of hexagonal fuel assemblies vibration considering mechanical nonlinearities

The paper presents a complex modelling concept for the nuclear fuel assembly vibration caused by the coolant pressure pulsations generated by the main circulation pumps. The method is based on the two-stage modelling with the reactor’s global model at the first stage and the detailed model of the particular fuel assembly at the second stage. This approach allows respecting all the relevant mechanical nonlinearities — clearances and friction in the key-groove couplings between the reactor core barrel and the pressure vessel and the internal impact or the friction interaction of the fuel assemblies. Numerical simulations allow better interpretation of monitoring data concerning the fuel burn-up.

Development of methods of gas chromatographic analysis of technological media of the main circulation pumps of a nuclear power plant

During the operation of power plant equipment, technological media are used – water, turbine oils. Diagnostics of equipment by chromatographic determination of diagnostic components (gases Н 2 , СН 4 , С 2 Н 6 , С 2 Н 4 , С 2 Н 2 , СО, СО 2 , О 2 , N 2 ), water, Ionol in these media is urgent. For this, 5 chromatographs are used. An increase in the reliability of the main circulation pumps of a nuclear power plant can be due to an increase in the reliability of its system. The influence of ultrasonic vibrations on the generation of gases in the systems oil – diagnostic gas, oil – water – diagnostic gas with the use of turbine Тп-22s is studied. Gas concentrations increase with an increase in the duration of exposure for 1600 s at a frequency of 35–125 kHz and a power of 20 W. The dependences of the concentrations Ci of dissolved gases on the irradiation time τ of technological media is expressed by the equation Сi=А∙τ+В Coefficients A, B and correlation coefficients R2 have specific values for each dissolved gas. So, 0.95≤R2≤0.995, which indicates the adequacy of the obtained equations to the experimental data. This makes it possible to determine turbine in water after irradiation and subsequent chromatographic determination of the generated dissolved gases. The technical requirements for a 5-channel chromatograph were established and its structural diagram was developed. This makes it possible to reduce the number of measurement operations and chromatographs. The thresholds for determining the diagnostic components in the corresponding technological environments have been determined: 2 ррm (Н 2 ); 1 ррm (СН 4 , С 2 Н 6 , С 2 Н 4 ); 0,5 ррm (С 2 Н 2 ); 5 ррm (СО, СО 2 ); 1,5 ppm (О 2 , N 2 ); 0.05 wt. % (Ionol); 2 g/t (water in turbine oil) 0.02 mg/dm3 (turbine in water). The basic technological scheme of the system for the main circulation pumps of the NPP has been developed. It is proposed to continuously: degassing the turbine flow; sorption purification of cooling water, analysis of turbine and cooling water by chromatography methods. This will reduce the degradation of turbine and increase the reliability of the system of the NPP main circulation pumps

Analysis of the connection of an idle MCP with three operating without a preliminary reduction in the power of the VVER-1000 reactor

The paper analyzes the transient process caused by connecting an idle loop (turning on an idle main circulation pump (MCP)) without preliminary reduction in the power of a VVER-1000 reactor plant (RP) operating at a rated power corresponding to three MCPs. The process under consideration is included in the “Reactive accidents” section of materials for justifying the safety of VVER-1000. Modeling is carried out on the example of the 3rd power unit of the Kalinin NPP [1]. For the calculations, we used the improved evaluation code ATHLET [2], which is included in the AC2 software package, officially obtained by the National Research Nuclear University MEPhI on the basis of a license agreement with Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS) gGmbH, Germany [3]. The ATHLET code has been certified in Russia for calculating stationary and transient conditions in water cooled reactors [4]. For the conservatism of the calculation, the study did not take into account the work of the PRL, APR, WP-1, WP-2 and AWP, the functioning of which is aimed at limiting the increase in the reactor power during the transient process. Blocking the signal to turn off the MCP on the fact of a decrease in the level in the SG and of the three ECCS systems, the operation of only one system is taken into account. For a number of reasons, this calculation was carried out for the end of the campaign. Implementation of design safety criteria for the adopted transition scenario is demonstrated.

The Development of the Technology for Producing Vessel Structures of the Main Circulation Pump for a BREST Reactor

Abstract—One of the joints of a liquid metal fast neutron reactor is the main circulation pump. The fabrication technology of the most labor-intensive joint of the main circulation pump, the pressure chamber, which is a complex welded structure made of 10Kh15N9S3B3 (EP-302Sh) sheet steel, is optimized and the fixtures and accessories are developed and fabricated, as a result of which the metal content of the structure is decreased. The process documentation for bending and stamping of the parts of the pressure chamber and welding of the components and their thermal and mechanical treatment and assembling is developed. The existing process documentation is adjusted as applied to the pressure chamber of the main circulation pump following the results of fabrication of this structure.

Experimental studies into the performance of the lead coolant axial pump wet ends to justify main circulation pumps for the HMLC reactor plant circuits

The paper presents the results of the studies to justify the design solutions for the main circulation pumps of the heavy liquid-metal cooled reactor plant circuits. A substantial difference has been shown in the performance of pumps for the heavy liquid-metal coolant transfer. The studies have confirmed the qualitative difference in the cavitation performance of coolants, the state of the gases and vapors they contain, the influence of supply and discharge devices, and the effects of the impeller blade section performance and geometry and the hub-tip ratio on the pump performance. The studies were performed based on NNSTU’s lead-cooled test facilities with the coolant temperature in a range of 440 to 550 °C and the coolant flow rate of up to 2000 t/h. The outer diameter of the impellers and the straightening devices was about 200 mm, and the thickness of the flat 08Kh18N10T-steel blades was 4.0 mm and that of the airfoil blades was up to 6.0 mm. The pump shaft speed changed in a stepped manner from 600 rpm to 1100 rpm after each 100 rpm. The studies were conducted to justify the engineering and design solutions for pumps as applied to conditions of small and medium plants with fast neutron lead cooled reactors currently under investigation at NNSTU (BRS-GPG). The experimental results can be recommended for use to design other HLMC transfer pumps.

Analysis of the Kalinin-3 coolant transient benchmark by SKETCH-N/SKAZKA code system

The switching off of the one of four main circulation pumps results in the Kalinin-3 Coolant Transient Benchmark. The benchmark contains four exercises, but this article shows the results of the exercise two of the VVER-1000 reactor core calculation. Reactor core analysis code SKETCH-N/SKAZKA is used to simulate the steady-state and transient conditions of the VVER reactor. The reactor core boundary conditions for the problem are provided by ATHLET/BIPR-VVER code.In the initial steady-state condition and at the end of the transient the power distribution is compared with the nuclear power plant (NPP) data. The average error in 3-D power distribution is 3.6%. The average difference in fuel assembly (FA) power is less than 2%. In the transient calculation, we compared with the experimental data the reactor power and axial offset. There is close agreement between the numerical results and NPP data.

Экспериментальные исследования характеристик проточных частей осевых насосов для свинцового теплоносителя в обоснование главных циркуляционных насосов контуров РУ с ТЖМТ

Экспериментальные исследования характеристик проточных частей осевых насосов для свинцового теплоносителя в обоснование главных циркуляционных насосов контуров РУ с ТЖМТ

Technology for producing vessel structures of the main circulation pump for BREST reactor

One of the important joints of a liquid metal fast reactor is the main circulation pump. In the course of the present research, a preproduction model of the pressure chamber, the most labor-consuming unit of the main circulation pump, was manufactured. This complex welded structure was made of 10Kh15N9S3B3 (EP-302Sh) steel sheets; fixtures and accessories were also manufactured, and as a result the metal consumption of the structure decreased. Technological documentation was made for bending, stamping of parts of the pressure chamber, welding of components, their heat and mechanic treatment, and its assembling. According to the results of this work, the existing technological documentation was adjusted in the context of the new data for pressure chamber of the main pump.

Simulation of a Transient Process in VVER-1200 by Means of the Athlet/BIPR-VVER Coupled Neutronics and Thermohydraulic Code

The results of a simulation, performed using the code ATHLET/BIPR-VVER (Russia, Germany), of tests on connecting a single main circulation pump (MCP) unit with three operating units during operation of VVER-1200 at 26% normal power level are presented. A comparison of the calculation and experiment showed the accuracy of the computational model to be satisfactory. Modeling is sufficient from the stand-point of calculating the macroparameters of a reactor facility.