Main Circulation Pumps Research Articles

Выбор запирающей среды уплотнения вала главных циркуляционных насосов РУ СО свинцовым и свинцово-висмутовым теплоносителями

Выбор запирающей среды уплотнения вала главных циркуляционных насосов РУ СО свинцовым и свинцово-висмутовым теплоносителями

New technology for purging the steam generators of nuclear power plants

A technology for removal of undissolved impurities from a horizontal steam generator using purge water is developed on the basis of a theoretical analysis. A purge with a maximal flow rate is drawn off from the zone with the highest accumulation of sludge in the lower part of the steam generator after the main circulation pump of the corresponding loop is shut off and the temperatures of the heat transfer medium at the inlet and outlet of the steam generator have equilibrated. An improved purge configuration is used for this technology; it employs shutoff and regulator valves, periodic purge lines separated by a cutoff fixture, and a D{sub y} 100 drain union as a connector for the periodic purge. Field tests show that the efficiency of this technology for sludge removal by purge water is several times that for the standard method.

Improvement of the equipment in fast-neutron reactor facilities

Information is presented on the development of the main equipment of the BN-1200 advanced reactor facility: first- and second-loop main circulation pumps, intermediate heat exchanger, actuation mechanism of the cold filter-trap control and protection system, autonomous and air heat exchangers, steam generator. The approach to the development of the equipment is based on maximum use of the experience gained in operating BN-350 and -600 as well as experience in developing the BN-800 design, which gives a basis for ensuring reliable operation of the BN-1200 equipment. New solutions for improving the technical-economic indicators and increasing the safety of a power-generating unit, whose validation required R&D work, are examined at the same time.

Large Break Loss-of-Coolant Accident Analysis of VVER-1000 Reactor Using CATHARE Code

In the safety analysis of nuclear power plants, large break loss-of-coolant accidents (LB LOCAs) continue to be one of the major issues. In this study, the latest version of the French best-estimate computer code CATHARE2 V2.5_1 Mod6.1 was used in order to predict the thermal-hydraulic phenomena in the VVER-1000 reactor during LB LOCA. This type of reactor is in operation and under construction in several countries including Russia, Ukraine, Bulgaria, Czech Republic, China, and India.The paper first presents the CATHARE modeling of a VVER-1000 reactor, including the core, the vessel, the primary and secondary circuits with the pressurizer, the main circulation pumps, the horizontal steam generators, and the steam lines. The emergency core cooling system (ECCS) is presented also with hydroaccumulators and high- and low-pressure safety injection pumps. The break, located in the cold leg and close to the reactor, is represented by the RUPTURE module of CATHARE, which modelizes a double-ended guillotine type of break.In the current calculations, the bottom-up and top-down reflooding play a very important role, compared to some western pressurized water reactors, because of the ECCS injection into the downcomer and upper plenum of the VVER-1000. That is why special attention is paid to the validation of the CATHARE reflooding model, based on the REWET-II experiment devoted to VVER. In CATHARE, a special two-dimensional reflooding module has been developed, which takes into account the radial and axial conduction in the vicinity of the quench front. The assessment based on the REWET-II facility shows the ability of the code to predict the progression of two simultaneous quench fronts.Then, the paper presents the results of the LB LOCA analysis for VVER-1000 (reference case and sensitivity study), describing the blowdown, refill, and reflood phases of the accident. Problems related to the countercurrent flow limitation (CCFL) phenomena at the core outlet are also considered. The results of the calculations show reasonable prediction of the basic parameters important to the safety of the plant, such as quench front progression, fuel-cladding temperatures, break flows, pressure behavior, etc. As expected, the temperatures—despite the conservative character of the modeling—remain below the safety criteria. It should be noted that the calculated fuel wall temperatures are very sensitive to the proper selection of the reflooding parameters and CCFL and need skillful choice of the reflooding conditions.

Design and analysis of a thermal hydraulic integral test facility for Bushehr nuclear power plant

In this paper, design and analysis of a thermal hydraulic integral test facility for Bushehr Nuclear Power Plant (NPP) is presented. The Bushehr Integral Test Facility (BITF) is a test facility designed to model the thermal-hydraulic behaviours of the Bushehr NPP (VVER-1000) pressurized water reactors currently in use in IRAN. These reactors have unique features that differ from other PWR designs. The BITF simulates the major components and systems of the reference NPP, making it possible to examine postulated small and medium break a loss of coolant accidents (LOCAs) and operational transients. The BITF is a volume-scaled model (1:1375). To ensure that gravitational forces remain equal to those in the reference reactor, the major components and systems in the BITF preserve 1:1 elevation equivalence to the reference reactor. The facility has four loops (each one consists of a hot leg, a steam generator, a loop seal, a main circulation pump and a cold leg), a pressurizer connected via two surge line to the hot leg of the loops 2, 4, the emergency-core-cooling system (ECCS) which is provided by an active pump simulating high and low pressure injection systems, and four hydro-accumulators. The report also contains a comparison between experimental data of PSB test facility and RELAP5 calculations of BITF facility under steady state condition of the reactor power 15% from the nominal.

Deterministic Analysis of Natural Circulation Events at the Ignalina NPP

Eight main circulation pumps (MCPs) are employed for the cooling of water forced circulation through the RBMK-1500 reactor at the Ignalina nuclear power plant (NPP). These pumps are joined into groups of four pumps each (three for normal operation and one on standby). In the case of all MCPs trip, the reactor shutdown system is activated due to decrease of coolant flow rate. At the same time, after the pump trip, the coolant to the reactor fuel channels during the first few seconds is supplied by pump coastdown. Later, the reactor is cooled by natural circulation. The main question arises whether this coolant flow rate is sufficient to remove the decay heat from the reactor core. This paper presents the investigation of all MCPs trip events at the Ignalina NPP by employing best estimate code RELAP5 and methodology of uncertainty and sensitivity analysis.

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.

Variable-frequency asynchronous drive used in circulation pumps of a power-generating plant

Features and engineering aspects associated with the development of variable-frequency electric drives are considered. The results of an investigation of their characteristics are presented. The electric drives are used in main circulation pumps that ensure the transportation of a liquid-metal coolant in a 25-MW nuclear power plant.

Investigation of the development of the 1986 accident at the Chernobyl nuclear power plant

Important results obtained from investigations of the diverse processes that occurred during the accident in the No. 4 unit of the Chernobyl nuclear power plant on April 26, 1986 are presented in a concise form. The events occurring in the unit in connection with the preparation for runout tests to be performed on a turbogenerator are examined: from the characteristics of the test program to the actions taken by the plant personnel in preparation for the these tests. The events that determined the character of the evolution and the catastrophic consequences of the accident are noted: involuntary stopping of the reactor and subsequent power increase without unpoisoning of the core, stopping of feed-water flow with increasing loop flow, operators’ lack of information about the actuation of the technological shielding of the electric drives of the main circulation pumps and electricity generator. Computational analysis of the initial phase of the accident and other computational assessments revealed new factors which with a large void effect made reactor runaway on prompt neutrons unavoidable-cavitation steam reaching the core entrance from the shutoff-regulation valve and/or the cavitation shutdown of the feed to the remaining main circulation pumps after the pumps powered by the run-out turbogenerator were disconnected. The most likely picture of the evolution of the accident is constructed on the basis of an examination of the destroyed power-generating unit.

Experience in the Operation of Adjustable Electric Drives of Main Circulation Pumps of BN-600 Reactors

Adjustable electric drives of the main circulation pumps of the BN-600 reactor of the Beloyarskaya NPP have a unique layout of asynchronous valve cascades (AVC). Twenty five years of successful operation of such drives prove the expediency of the use of AVC, which deserves study and application in the design of power units with BN-800 reactors.

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

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

Optimization of Conceptual Solutions for the RBETs-M Lead–Bismuth Fast Reactor

An improved design for the RBETs-M reactor is examined. The design is characterized by innovative solutions to various problems. These solutions make it possible to improve safety and cost-effectiveness as compared with the base RBETs design. The innovative solutions include using 15N-based nitride fuel, a two-loop scheme with no main circulation pumps, a gas-lift system in the first loop, passive systems for removing heat and stopping the reactor, and others.

Investigation of the Mixing of Coolant Flows in a VVÉR Vessel

The mixing of flows in a flow-through channel of a VVER-1000 reactor in the case where a flow with high or low boron concentration and a flow whose temperature is different from the temperature in the reactor are fed into the same circuit is studied. To this end investigations were performed on a bench at the Special Design Office Gidropress with a model VVER-1000 reactor on a 1:5 scale. Regimes with the main circulation pump switched on and with natural circulation restored for an accident with a small leak are studied.

Development of 160 m3/min Large Capacity Sodium-Immersed Self-Cooled Electromagnetic Pump

A large capacity sodium-immersed self-cooled electromagnetic pump (LEMP) was developed for application to the main circulation pumps of FBR. This advanced LEMP is a submergible annular linear induction pump designed to be self-cooled by immersing into sodium and applying high temperature electrical insulation. Almost all the internal electrical losses were transferred to the surrounding sodium, which can be recovered as electricity by turbine generators. The LEMP having specifications of 160m3/min flow rate, 0.28 MPa head and more than 40% efficiency at the rating was designed, fabricated and tested in the sodium pump test facility. The test involves magnetic field measurement in the air and a variety of sodium tests during 2,550 h, which demonstrated good pump performance and flow controllability, and satisfied the design target. The boundary between flow stability and instability of the LEMP operation could be defined by peak position of the Q-H curve, which was specified by Rem×S (magnetic Reynolds' number times slip) of 1.4 to 1.5 at 335°C. Based on the test results, the applicability of the LEMP for the FBR was confirmed.

Quantitative assessment of MCP trip transient in a VVER

Pressurized water vessel-type reactor (VVER) safety has become a very important issue, in particular for countries in Central and Eastern Europe. For thermal-hydraulic analyses the western codes like RELAP5, CATHARE and ATHLET were used. The purpose of the study was to quantitatively assess the RELAP5 capability to predict the main circulation pump (MCP) trip at nearly full power transient in Mochovce VVER 440/213 nuclear power plant (NPP). The transient parameters were recorded during the start up test program implementation. For accuracy quantification the improved fast Fourier transform based method (FFTBM) was used. The RELAP5/MOD3.2.2 computer code was used for calculation. The results showed very good agreement between calculated and plant measured data. The results also confirmed some previous studies that the simpler is the transient the higher code accuracy is generally achieved.

Uncertainty and sensitivity analysis of MCPs’ trip events at Ignalina NPP

Eight main circulation pumps (MCPs) are employed for the cooling water forced circulation through the RBMK-1500 reactor at the Ignalina nuclear power plant (NPP). There have been a few events when one or more MCPs were inadvertently tripped. This paper presents investigation of a one MCP trip event and all MCPs’ trip events at Ignalina NPP. Thermal-hydraulic analysis was conducted using the best estimate system code RELAP5/MOD3.3. Uncertainty and sensitivity analysis of flow energy loss in different parts of the main circulation circuit (MCC), initial conditions and code-selected models was performed. Such analysis allows to estimate the influence of separate parameters on the calculation results and find those modelling parameters that have the largest impact on the investigated events. Uncertainty analysis indicates that natural circulation provides adequate cooling in the case of all MCPs tripped, and that the reactor is reliably cooled by forced circulation in the case of a single tripped MCP. On the basis of this analysis, recommendations for the further improvement of model are developed.

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

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

Benchmark analysis of main circulation pump trip events at the Ignalina NPP using RELAP5 code

The state-of-the-art code RELAP5/MOD3 was originally designed for PWRs. Because of unique RBMK designs the application of this code to RBMK-1500 encountered several problems. A successful best estimate RELAP5 model of the Ignalina NPP has been developed. This model includes the reactor main circulation circuit (MCC) and reactor control and protection system required for this kind of transient analysis. Benchmark analysis of all operating main circulation pump (MCP) trip events was performed. During the analysis the characteristics of isolation control valves and MCP throttling regulating valves were established. Comparison of calculated and measured parameters was also used to establish realistic resistances of different MCC components and realistic behaviour of the controllers of the reactor systems. Calculations performed with the RELAP5 model, which includes these modifications, compare favourably with plant data.

Development of a Computer Code, ONCESG, for the Thermal-Hydraulic Design of a Once-Through Steam Generator

Development of the conceptual design of a 300 MWt integral reactor, SMART (System-integrated Modular Advanced ReacTor), for utilization in nuclear cogeneration plants has been completed at the Korea Atomic Energy Research Institute (KAERI). The major primary components of the SMART such as modular helical steam generators, main circulation pumps and a self regulating pressurizer are integrated into a reactor vessel. It is a common practice to employ a once-through steam generator in integral reactor designs because of its advantages in compactness and simplicity of the flow path arrangements. In this study, a thermal hydraulic design and performance analysis computer code for a once-through steam generator using helically coiled tubes, ONCESG, is developed. To benchmark the developed physical models and computer code, once-through steam generators developed by other designers are simulated and ONCESG calculated results are compared with the design data. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by ONCESG showed general agreements with the published data and it is demonstrated that the ONCESG code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

A simple nonlinear model for pendular chaotic oscillations of reactor core barrel

We propose a non-linear model reproducing the unidirectional (quasi)pendular motion of pressurized water reactor (PWR) core barrel (CB), which is in intermittent contact with reactor pressure vessel (RPV) in places of lower fastenings. Analytic formulae are obtained, which allow one to elucidate the influence of reactor parameters on the character of motion. We show by the example of the WWER-1000 reactor that, in the framework of our model, there exists a possibility of chaotic pendular oscillations in the certain region of three parameters, namely, (i) the pendular oscillation frequency, (ii) the resulting force acting on the CB at the main circulation pump (MCP) frequency from the coolant flows of opposite legs, and (iii) the clearance between the CB and RPV. Vibratory monitoring, which is realized in modern diagnostic systems, in principle, allows one to distinguish between regular and chaotic regimes, thus giving information about the state of fastenings in the lower part of the CB and RPV. We also demonstrate the possibility of stabilizing chaotic oscillations by applying recently proposed control technique.