Liquid Metal Coolant Research Articles

ПОДДЕРЖАНИЕ КАЧЕСТВА СВИНЕЦСОДЕРЖАЩИХ ТЕПЛОНОСИТЕЛЕЙ В ПЕРСПЕКТИВНЫХ ТЕХНОЛОГИЯХ ПЕРЕРАБОТКИ ОРГАНИЧЕСКОГО И НЕОРГАНИЧЕСКОГО СЫРЬЯ

The unique properties of heavy liquid metal coolants (low chemical activity when interacting with air, organic compounds, water and water vapor, high boiling point, low vapor pressure) allow them to be used both in the first circuits of fast neutron reactor plants and in heat exchangers used for processing various organic and inorganic raw materials. Such devices can be either recuperative or mixing type. The latter additionally have the following advantages: lower cost and simplicity of construction; no heat transfer surfaces that can be subject to corrosion and contamination; reduced dimensions due to a larger specific heat transfer surface. The greatest degree of study relates to liquid metal pyrolysis of solid substances, direct contact distillation of aqueous solutions without preliminary preparation, improving the efficiency of continuous casting machines due to the replacement of water with lead-bismuth eutectic. The results of research in support of the above processes are presented. The methods of maintaining the quality of lead-containing heat carriers do not have fundamental differences for the conditions of reactor installations and in relation to promising technologies for processing organic and inorganic raw materials. At the same time, the requirements for devices for implementing these methods in promising technologies for processing various raw materials are significantly lower: in general, a sensor for measuring the thermodynamic activity of oxygen and a device for entering gas (hydrogen-containing or oxygen-containing) mixtures in the coolants.

Liquid metal MHD flow influence on heat transfer phenomena in fusion reactor blankets

DEMO reactor is projected as a technological demonstrator and a component test reactor for several key systems such as the breeding blanket. The dual coolant lithium lead (DCLL) is one promising breeding blanket candidate that performs the functions of shielding, part of the cooling and breeding with the same liquid metal fluid. PbLi is also in charge of transporting the bred tritium to the Tritium Extraction System. Structural cooling of the blanket walls is performed by pressurized helium flow which also transfers heat to the balance of plant. Keeping a proper distribution of thermal loads between the two coolants is a key aspect to reach a plant high efficiency. The thermal power transferred from the liquid metal to the helium channels depends on the flow regime. While liquid metal flows under the magnetic field, the magnetohydrodynamic (MHD) phenomena will occur, influencing the flow regime and the velocity profile. The heat is volumetrically deposited in the liquid metal coolant due to the interaction between the highly energetic neutrons originated in the plasma and the coolant. The heat deposition profile is roughly exponential causing temperature differences across the same channel cross section. Temperature gradients induce buoyant effects altering the expected MHD profile and affecting the heat transfer rate from the liquid metal to the surrounding walls cooled with helium. It is of utmost interest to identify the effects of the main flow parameters on the heat transfer coefficient under such complex conditions. The work studies computationally a domain consisting of a liquid region and a surrounding solid region electrically and thermally coupled using latest EU DCLL design characteristics. A fully developed flow is assumed where buoyancy is modelled under the Boussinesq approximation. Joule effect has been considered negligible compared to the neutron deposition. The influence of the buoyant MHD phenomena to the heat transfer coefficient is analysed parametrically with a CFD solver implemented in OpenFOAM®. The parameters investigated here are the Reynolds, Hartmann and Grashof numbers, the Grashof ratio, and the wall conductance ratio (cw).

Development of a methodological approach for the computational investigation of the coolant flow in the process of the sodium cooled reactor cooldown

A methodological approach has been developed for the computational investigation of the thermal-hydraulic processes taking place in a sodium cooled fast neutron reactor based on a Russian computational fluid dynamics code, FlowVision. The approach takes into account the integral layout of the reactor primary circuit equipment and the peculiarities of heat exchange in the liquid metal coolant, and makes it possible to model, using well-defined simplifications, the heat and mass exchange in the process of the coolant flowing through the reactor core, and the reactor heat-exchange equipment. Specifically, the methodological approach can be used for justification of safety during the reactor cooldown, as well as for other computational studies which require simulation of the integral reactor core and heat-exchange equipment. The paper presents a brief overview of the methodological approaches developed earlier to study the liquid metal cooled reactor cooldown processes. General principles of these approaches, as well as their advantages and drawbacks have been identified. A three-dimensional computational model of an advanced reactor has been developed, including one heat-exchange loop (a fourth part of the reactor). It has been demonstrated that the FlowVision gap model can be applied to model the space between the reactor core fuel assemblies (interwrapper space), and a porous skeleton model can be used to model the reactor’s heat-exchange equipment. It has been shown that the developed methodological approach is applicable to solving problems of the coolant flow in different operating modes of liquid metal cooled reactor facilities.

A novel liquid lithium jet-cooled finger-type divertor target concept for fusion power plant application

The divertor target is the most thermally loaded plasma-facing component in a foreseen DEMO reactor and beyond, which has to tolerate the peak high heat fluxes of up to ∼20 MW m−2 produced by intense plasma bombardment, radiation and nuclear heating. However, none of current designs including water-cooled and helium-cooled concepts can satisfy this requirement. Motivated by the excellent power removal capacity of liquid metal coolant and combined with the structure characteristics of the finger-type helium-cooled target, a novel concept of liquid Li jet-cooled finger-type divertor target for DEMO reactors was proposed in this paper. The performance analysis, including thermal-hydraulics analysis, mechanical analysis and MHD effects analysis, have shown that the proposed design can withstand 20 MW m−2 heat load because the temperatures of the structural materials remain within the thermal rules and the maximum thermo-mechanical stress in the VM-W thimble is approximately 484 MPa appearing in the round corner, which is below the 3S m limit at the corresponding temperature. Moreover, a theoretical and empirical analysis has confirmed that MHD effects on pressure drop and heat transfer is rather limited in the design. The comparison of this new design with other representative designs including water-cooled ITER-like target design and helium-cooled modular jet target design has been made, and the results shows that the proposed liquid Li cooled target design has better performance under 20 MW m−2 high heat flux and ∼10 dpa neutron irradiation. Therefore, this design is promising to provide a new option for solving the DEMO reactor divertor heat removal issues. Certainly, a large number of R&D efforts are still needed to ensure the success of this concept, particularly in the areas of materials, fabrication and irradiation.

Health risk analysis as per radiation and chemical factors in a zone influenced by a nuclear legacy object in the north-western part of Russia

Rehabilitation programs for nuclear legacy objects in north-western regions in Russia involve decommissioning; a former coast technical support base belonging to the RF Navy and located on the Gremikha peninsula in Murmansk region is now being removed from service. At present spent nuclear fuel is being extracted from ship reactors with liquid-metal coolant and nuclear wastes that have been accumulated during active operations are being treated. The current activities at the site result in complex influence on the environment determined by both radiation and non-radiation contaminants. The present work focuses on examining a hygienic situation existing in residential areas located close to this nuclear legacy object. In general, radiation situation in residential areas is determined by technogenic contaminants, basically 137Cs and 90Sr radionuclides. Our research has revealed that at the moment their contents in the environmental objects correspond to background levels that are typical for the examined region. Chemical contamination detected in residential areas is characterized with elevated heavy metals contents (Cd, As, Pb, V, Zn, Cu etc.) in soils in urban settlements that are higher than background levels. We detected lead and zinc concentrations that were higher than MPC. Total soil contamination with heavy metals (Zc) is still «permissible» in most residential areas. Carcinogenic health risk for people living in the examined residential areas does not exceed 10-6 as per radiation and chemical factors. Research results allow assessing the current hygienic situation as normal and concluding that decommissioning of SZC «SevRAO» objects (Gremikha section) does not exert negative impacts on the environment and health of people living in residential areas located nearby. The obtained data give grounds for controlling activities when environmental monitoring is accomplished during works aimed at recovering contaminated areas where nuclear legacy objects are located.

Анализ риска по радиационным и химическим факторам для населения в районе расположения объекта ядерного наследия в северо-западном регионе России

Rehabilitation programs for nuclear legacy objects in north-western regions in Russia involve decommissioning; a former coast technical support base belonging to the RF Navy and located on the Gremikha peninsula in Murmansk region is now being removed from service. At present spent nuclear fuel is being extracted from ship reactors with liquid-metal coolant and nuclear wastes that have been accumulated during active operations are being treated. The current activities at the site result in complex influence on the environment determined by both radiation and non-radiation contaminants. The present work focuses on examining a hygienic situation existing in residential areas located close to this nuclear legacy object. In general, radiation situation in residential areas is determined by technogenic contaminants, basically 137Cs and 90Sr radionuclides. Our research has revealed that at the moment their contents in the environmental objects correspond to background levels that are typical for the examined region. Chemical contamination detected in residential areas is characterized with elevated heavy metals contents (Cd, As, Pb, V, Zn, Cu etc.) in soils in urban settlements that are higher than background levels. We detected lead and zinc concentrations that were higher than MPC. Total soil contamination with heavy metals (Zc) is still «permissible» in most residential areas. Carcinogenic health risk for people living in the examined residential areas does not exceed 10-6 as per radiation and chemical factors. Research results allow assessing the current hygienic situation as normal and concluding that decommissioning of SZC «SevRAO» objects (Gremikha section) does not exert negative impacts on the environment and health of people living in residential areas located nearby. The obtained data give grounds for controlling activities when environmental monitoring is accomplished during works aimed at recovering contaminated areas where nuclear legacy objects are located.

Calculation of flow heat transfer characteristics of liquid gallium

In order to design a more compact small nuclear reactor, liquid metal gallium can be selected as the reactor coolant. Gallium is rarely used in reactor as liquid metal coolant, however, due to its excellent thermal conductivity and low melting point, research on the flow and heat transfer characteristics of liquid metal gallium in channels can provide a theory basis for nuclear reactor design optimization and relax reactor safety limits. The flow heat transfer characteristics of liquid gallium are obtained by calculating the surface heat transfer coefficient of liquid gallium in various thermal and structural parameters. The calculation results show that the heat transfer between liquid gallium and the pipe wall can be enhanced by the increase of the inlet flow velocity, the increase of the wall heat flux density, or the reduction of the pipe size. When the hydraulic diameter of each pipe is the same, the heat exchange effect of liquid gallium flow in tubes is: triangle tube>square tube>round tube. The flow heat exchange effect of liquid gallium is better than that of lead-bismuth fluid.

THE UK NATIONAL THERMAL-HYDRAULICS FACILITY: MOTIVATIONS, DESIGN AND PLANNING STATUS

A UK National Thermal-Hydraulics Facility (NTHF) dedicated to supporting new reactor and other relevant business is being developed, one of the purposes being to deliver on the government’s carbon emission reduction commitments. The facility site is foreseen to be at Menai Science Park on the isle of Anglesey in North Wales, a region expected to see significant low carbon energy deployment in coming years. The UK NTHF is envisioned to cater for the needs of emerging nuclear in the UK – but also to serve as a hardware platform for international thermal-hydraulics research collaboration. Plans are to construct a platform capable of maintaining several test loops including support for the UK’s on-going, conventional nuclear new build programme as well as Gen-IV systems and associated materials like molten salt and liquid metal coolant media. Motivations are given for NTHF expected capabilities and requirements, which form the basis for its current design and planning state.

Modeling of Gas-Phase Transport in Heavy Liquid-Metal Coolant Flow in the HYDRA-IBRAE/LM Thermohydraulic Code

Approaches to modeling gas-phase transport in heavy liquid-metal coolant flow in the HYDRA-IBRAE/LM code, which is actual for safety validation of nuclear reactors with a leak in the steam-generator tubes, are examined. A hierarchy of models for determining the dispersed-particle size that are implemented in the code for bubble flow is presented: empirical correlation, transport equation for the density of the interfacial surface area, and kinetic equation for the particle-size distribution function. The results of test calculations are presented.

ОЦЕНКА НИЗКОКИСЛОРОДНОЙ ГРАНИЦЫ СУЩЕСТВОВАНИЯ ПЛЮМБОФЕРРИТНЫХ ФАЗ В РАСПЛАВЕ СВИНЦА

At present, installations with heavy liquid metal coolants (HLMC) based on lead melts are considered as promising nuclear power plants, since these coolants have a number of advantages over alkali metals and other coolants. A feature of heavy liquid metals is their rather high corrosivity towards structural materials; however, at present this problem has been partially solved due to the formation of protective oxide coatings on the surfaces of steels based on oxides of metal components of structural steels. This determines the main tasks of modern HLMC technology, which are to provide conditions for the formation and maintenance of oxide layers with optimal protective properties on the surfaces of structural steels. The success of solving these problems in the general case depends on an adequate representation and quantitative description of the properties of the system “liquid metal melt - oxide phase”. In this work, we consider the questions of the possibility of the formation of plumboferrite phases in liquid lead, their composition and thermodynamic properties.

Radionuclide Chemistry in Nuclear Facilities Based on Heavy Liquid Metal Coolants: Past, Present and Future.

Heavy liquid metals such as lead and lead bismuth eutectic (LBE) are considered as spallation target material for next-generation neutron sources and as coolant of fast spectrum nuclear reactors that are developed to facilitate more efficient use of nuclear fuel as well as transmutation of long-lived nuclear waste. During the operation of such facilities, the heavy liquid metal will be activated by nuclear reactions. Additionally, fission product radionuclides may be introduced into the liquid metal from leaking fuel pins or by fission of the target nuclei in spallation. The chemical behaviour of these radioactive contaminants in the liquid metal - especially their immediate volatilization or volatilization of formed secondary compounds - may affect the safety of such facilities. The present article summarizes the activities of PSI's Laboratory of Radiochemistry towards a better understanding of the chemistry of potentially hazardous radionuclides in LBE and discusses aspects that need to be addressed in future to support the licensing of heavy liquid metal-based nuclear facilities.

Conceptual Design of a Hybrid Fusion–Fission Power Reactor with a Tokamak and a Depleted Uranium Blanket

We assess the viability of energy generation with depleted uranium irradiated with fast neutrons from the reaction of DT fusion driven by a tokamak reactor. Since existing nuclear power plants employ uranium fuel enriched to 4.5% 235U, over 90% of natural uranium currently goes to nuclear waste and is no longer used for power generation. The amount of depleted uranium with the 235U content of 0.2–0.3% accumulated in the world exceeds 1 500 000 t and is continually increasing owing to the operation of thermal neutron power plants. Industrial fast neutron power plants using depleted uranium as nuclear fuel are still at the development stage. Therefore, alternative approaches to utilizing depleted uranium for power generation should be considered. Compared to fast reactors with a closed cycle utilizing plutonium, the proposed facility involves less radiation hazard since the irradiated fuel has to be reprocessed and reloaded less frequently. Apart from that, the proposed hybrid power reactor operates in the deeply subcritical regime, which provides a high level of nuclear safety. A tokamak reactor similar to that of the ITER project is selected as a source of a high-power flux of fusion neutrons whereby a reactor blanket of depleted uranium is irradiated. The heat will be removed with a liquid metal coolant rather than with water as in the ITER reactor. This will help maintain the hardness of the neutron spectrum, thereby maximizing the rate of 238U fission. With such a neutron source and a blanket of natural uranium with an optimized thickness of ~20 cm, each fission of uranium (with an energy release of 200 MeV) is estimated to finally produce four 239Pu atoms in the blanket (see the 2009 proceedings of the ROSATOM commission on optimizing the development of power generation with tokamaks). This estimate equally applies to a reactor blanket of depleted uranium. According to estimates, an “ideal” reactor with a blanket of depleted uranium operating in a stationary mode with a capacity factor of 70% would generate nearly 1.5 GW of electric power. However, only an estimated one-third of the total power can actually be delivered to the electrical network since the rest is required for breeding tritium and powering the facility itself. Toward maximizing the capacity factor, the first wall of the fusion reactor will be coated with lithium for protection against plasma-induced erosion.

Integrated researches of topical problems of fast reactors thermal physics

The results of actual studies are presented to justify the thermophysical characteristics and safety of fast reactors with sodium and lead coolants. The influence of a stratified coolant flow on the formation of velocity and temperature fields in a fast reactor vessel with a sodium coolant is shown. The development of the degradation of the fuel assembly during the development of a severe accident with the loss of sodium flow rate in a fast reactor, the blockage of the cross sections of the fuel assembly and the release of marker materials beyond it are demonstrated. The expediency of the combined sodium purification system built into the reactor vessel is shown, in which “cold traps” are an indispensable element, and “hot traps” provide accelerated oxygen purification during operation of the NPP in nominal regime. The results of studies of a large-modular sodium-water steam are presented. As applied to a reactor with a lead coolant, the influence of spacer grids on heat transfer in a fuel assembly of the core is shown. On the model of the steam generator of the reactor installation with lead coolant, it was found that the values of the steam temperature at the outlet of both collectors coincide. The temperature of lead at the exit from the lowering section and in the main path of lead is coincides also. The state and prospects of the development of technology for heavy liquid metal coolants are analyzed. The principle possibility of providing the required parameters of a high-temperature fast reactor with a sodium coolant for the production of hydrogen is shown. The problems of further thermophysical research are analyzed.

New Version of the KTM Lithium Divertor

At present, the Kazakhstan Tokamak for Material testing (KTM) project is in progress. The optimization of innovative design elements of the divertor is carried out at the KTM, in addition to studying materials applied for fusion technologies. The design of the lithium divertor module based on lithium capillary porous systems has been developed and tested successfully in real tokamak conditions. In this design, the problem of removing high-density heat flows is solved by using liquid metal coolant based on the Na–K eutectic alloy. The requirements for the improved safety and the divertor design compatibility with other in-vessel water-cooled elements of the tokamak and temperature limitation of the receiving lithium surface to the level <600°C at the heat flows of 10–20 MW/m2 stipulated a new design solution of the divertor experimental module and the use of a totally new coolant—water flow dispersed with gas (gas-water spray). In the paper, the design solutions of the new version of the module, the coolant parameters, and the cooling system layout are described and substantiated. The experimental results on determining the heat transfer coefficient of the coolant based on gas-water spray are considered.

Thermodynamics of equilibrium states and approaches to analyzing the mass transport in metal-oxide systems

Analysis of corrosion processes has a major role in justifying the reliability and safety of developed nuclear reactors of a new generation with heavy liquid metal coolants. An approach has been developed which allows practical conclusions to be made with respect to the processes in the given metal-oxide system based on analyzing state diagrams for these systems in the "oxidation potential – temperature" coordinates. The proposed approach relies on a long-term experience of experimental and computational studies concerned with the interaction of various steel grades with molten lead and lead-bismuth, as well as with the transport of metal impurities within these molten metals. The oxidation potential of a metal-oxide system is measured in experimental studies using oxygen activity sensors developed and manufactured at IPPE. The applicability of the proposed approach to analyzing the processes of mass transport in iron-oxygen, lead-oxygen, sodium-oxygen, and iron-water vapor systems has been demonstrated.

Basic models and approximation for the engineering description of the kinetics of the oxide layer of steel in a flow of heavy liquid metal coolant under various oxygen conditions

The article presents the results of corrosion processes, kinetics and changes in the oxide layer modeling using MASKA-LM software complex. The complex is intended for a numerical simulation of three-dimensional non-stationary processes of mass transfer and interaction of impurity components in a heavy liquid metal coolant (HLMC: lead, lead-bismuth). The software complex is based on the numerical solution of coupled three-dimensional equations of hydrodynamics, heat transfer, formation and convective-diffusive transport of chemically interacting components of impurities. Examples of calculations of mass transfer processes and interaction of impurity components in HLMC, formation of protective oxide films on the surfaces of steels are given to justify the coolant technology.

Codes of new generation for safety justification of power units with a closed nuclear fuel cycle developed for the “PRORYV” project

The article describes the status of development of codes of new generation for the “PRORYV” Project by the end of 2019: twenty-five commercial-grade software products to justify design solutions and safety of power units with fast neutron reactors and liquid metal coolant (sodium and lead) in a closed nuclear fuel cycle. The developed system of codes is multi-physical and multi-scale that allows performing both calculations of the whole installations and high precision calculations of their individual elements. The developed codes offer unique features. Twelve developed codes have already been certified by Rostechnadzor, and six more have been submitted for certification. In addition to creating the software products, a large-scale work is being carried out to conduct experimental studies for code validation that meet modern requirements imposed by the codes: unique measurement techniques have been created; experimental data on flow characteristics of heavy liquid metal coolant (HLMC) in a fuel assembly simulator have been obtained, as well as of “gas-HLMC” interphase interaction after inert gas injection in HLMC and characteristics of heat exchange between the inert gas and HLMC. The results are already used for validation of system and CFD codes used in the “PRORYV” Project.

Numerical and analytical modelling of concurrent physico-chemical processes passing at the boundary of steel in a liquid-metal coolant leading to film formation

We analyse an engineering model for a self-consistent calculation of the growth of an oxide film in circulation circuits with a heavy liquid metal coolant and concentrations of impurities (oxygen, iron, magnetite) from the point of view of possible uncertainty in determining the activity of oxygen. The modelling of thermohydraulic and physicochemical processes is based on solving the associated three-dimensional equations of hydrodynamics, convective-diffusive transport, and the formation of chemically interacting impurity components in the coolant volume and on the surface of steels. Model calculations of the influence of the uncertainty of oxygen activity on the rate and integral yield of iron, which under the given conditions of the oxygen regime after interaction with oxygen determines the appearance of magnetite. It is numerically demonstrated that in the saturation mode there is a model-independent characteristic, which is determined by the parabolic constant and thickness of the oxide film characteristic of steel. Understanding these processes is highly important for justification of loops with HLMC and steadiness of structural materials that are in contacts with HLMC.

Research on slipper bearings of pumps of reactor systems with lead-bismuth coolant at relatively low temperatures

The academic focus of the article relates to nuclear power engineering, reactor plants (RP) with heavy liquid metal coolant (HLMC) – lead and eutectic lead-bismuth. The article’s purpose is to set out the goals and objectives of the research of slipper bearings for HLMC transfer pumps and to identify the problems of design and operation under these conditions based on the available experience. This experience applies both to energy reactors with high HLMC temperatures (400-550°C), and partly to acceleration-controlled systems and research reactors with HLMC temperature of about 200-350°C. The choice of the type of slippper bearing is justified for further researches of its operability in conditions of low HLMC temperatures.

Preparation and Characterization of Nb-1Zr-0.1C Alloy Suitable for Liquid Metal Coolant Channels of High Temperature Reactors

Abstract A novel process comprising of aluminothermic coreduction of mixed oxides followed by arc and electron beam melt refining was developed for preparation of Nb-1Zr-0.1C alloy. The parameters of the process were optimized by considering the thermodynamic (heat) and mass balance phenomenon. The ingots of the homogenized alloy produced after electron beam melt consolidation were further extruded into tubes. The alloy was vacuum annealed at 1350–1800 °C to study the stability of Nb2C and Nb(Zr)C carbide precipitates in the microstructure. Compression creep tests conducted at 900 and 1000 °C revealed a stress exponent value of 2 and activation energy of 508 kJ/mol. NbSi2-based coatings were developed on the Nb-1Zr-0.1C alloy tubes using pack siliconizing process. The coated alloy was tested for oxidation at 1250 °C, and corrosion in liquid lead-bismuth eutectic (LBE) alloy at 875 °C for prolonged duration. The silicide-coated alloy showed superior oxidation and LBE corrosion resistance at high temperatures. The alloy was found to be a promising material for coolant channels of high temperature reactors.