Water-cooled Reactor Research Articles

Thermal-hydraulic behavior simulations of the reactor cavity cooling system (RCCS) experimental facility using Flownex

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

Results and lessons learned from the Generation IV SCWR-FQT comprehensive Monte Carlo computational benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor technology has been in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. A specific work package has been dedicated to studying the design- and safety-related neutronic parameters and reactor physics behavior of the SCW-SMR to support the pre-conceptual design process. Three Monte Carlo codes, viz., MCNP, OpenMC, and Serpent, were selected for pre-conceptual design applications and code-to-code comparison within the Gen-IV SCWR-FQT reactor physics computational benchmark. The effective multiplication factor, the axial power distribution within the fuel, the axial three-group neutron flux distribution, and the spatial distribution of the energy deposition due to neutron and photon interactions were determined. In this paper, results and lessons learned from this study are presented, and useful considerations are summarized to provide guidance in obtaining consistent results among the three Monte Carlo codes.

A water cooled, high power, dielectric barrier discharge reactor for CO2 plasma dissociation and valorization studies

Aiming at the energy efficient use and valorization of carbon dioxide in the framework of decarbonization studies and hydrogen research, a novel dielectric barrier discharge (DBD) reactor has been designed, constructed and developed. This test rig with water cooled electrodes is capable of a plasma power tunable in a wide range from 20W to 2 kW per unit. The reactor was designed to be ready for catalysts and membrane integration aiming at a broad range plasma conditions and processes, including low to moderate high pressures (0.05–2 bar). In this paper, preliminary studies on the highly endothermic dissociation of CO2, into O2 and CO, in a pure, inert, and noble gas mixture flow are presented. These initial experiments were performed in a geometry with a 3 mm plasma gap in a chamber volume of 40cm3, where the process pressure was varied from few 200 mbar to 1 bar, using pure CO2, and diluted in N2. Initial results confirmed the well-known trade-off between conversion rate (up to 60%) and energy efficiency (up to 35%) into the dissociation products, as measured downstream of the reactor system. Improving conversion rate, energy efficiency and the trade-off curve can be further accomplished by tuning the plasma operating parameters (e.g. the gas flow and system geometry). It was found that the combination of a high-power, water-cooled plasma reactor, together with electronic and waveform diagnostic, optical emission and mass spectroscopies provides a convenient experimental framework for studies on the chemical storage of fast electric power transients and surges.

Nuclear reactor Thermal-Hydraulics Special and prospective topical areas

The purpose of this paper is to summarize the session on Special & Prospective Areas (SPA) as part of the OECD/NEA/CSNI Specialists Meeting on Transient Thermal-hydraulics in Water Cooled Nuclear Reactors (SM-TH). This session dealt with some special and prospective topical areas in current nuclear reactor thermal hydraulics research. These topics are at the intersection of nuclear reactor thermal hydraulics in its traditional sense and multi-scale, multi-physicochemical phenomena associated with a nuclear power plant. The intricacy of nuclear power plants and the variety of possible accident conditions point to the complexity and need for an interdisciplinary approach to assess those phenomena. In addition to the nuclear or power technology itself, a range of unpredictable human actions introduce additional uncertainties in an already complex environment and the events’ initiation and propagation, which may indicate our inability to cover some unexpected or poorly understood phenomena. This summary should not be construed as conclusive but rather as a starting point and invitation for further discussions and exploration of gaps in our understanding of the phenomena involved.

Cooldown of a water-cooled reactor during the natural circulation mode using decay heat of the core and a low-power steam turbine

The Fukushima-1 nuclear power plant accident in 2011 confirms that NPP safety is one of the most urgent present-day problems. Moreover, a rapidly growing number of NPP units increases the risk of serious radiation accidents. The authors have developed a novel cooldown method for water-cooled reactors to be used under loss of connection with the power system using decay heat from the core and a low-power steam turbine. A distinctive feature of the method is financial viability of the steam turbine due to its possibility to generate additional electricity at the normal conditions of NPP operation. The aim of the invention is to ensure a high level of NPP safety while maintaining and increasing commercial efficiency and competitiveness of nuclear power plants. To implement the developed method, it is necessary to provide guarantee for reliability of NPP units at natural circulation of the primary coolant, since the primary circulation pumps will be shut off caused by a trip of a reactor. For this reason, the authors carried out an investigation into international practice theory and actual usage of natural circulation of water-cooled reactors. Experiments conducted at Kozloduy NPP and research of a number of scientists demonstrate a possibility to maintain constant decay heat removal that will ensure obtaining steam with stable parameters in the secondary circuit. The second part of the research deals with a new method aimed at beneficial usage of steam obtained by using the decay heat from the core under natural circulation of the primary coolant. The generated steam can be used in the additional low-power steam turbine to power supply the NPP's auxiliary needs. In our earlier research we performed a preliminary assessment of reliability of the developed system. The results of the research showed that a combination of a steam turbine and a standard three-channel emergency power supply system with diesel generators enables a safety level comparable to that of external heat exchangers in the passive heat removal systems. At the same time, passive heat removal systems require significant capital investments and maintenance costs, whereas according to the method developed by the authors, a low-power steam turbine can operate in the normal regime to generate additional electricity, as well as in combination with energy storage systems. This will ensure full cost recovery of the developed installations. To assess a reduction of risks of nuclear accidents in terms of cash equivalents when using the novel method, the authors analyzed the influence of various factors on damage expectancy in case of an accident at NPP, based on the example of Fukushima and Chernobyl NPPs. The investigation showed that a reduction of nuclear accident risk in terms of cash equivalents, when implementing the authors’ method aimed to supply the auxiliary NPP needs in blackout situations, reaches tens of millions of US dollars per year.

Integral and Separate Effects Test Facilities To Support Water Cooled Small Modular Reactors: A Review

This study reviews previous experimental facilities and test programs relevant to water-cooled reactor system design and analysis to meet regulatory compliances. This study aims to find the best solution for designing the required experiments, obtaining necessary test data, and verifying the developed computer code/models to support the new reactor design and development while minimizing cost and time while leveraging experiences from previous facilities to minimize. Nuclear reactor licensing requires supportive design, analysis, and experimental results to ensure the safety of the full-scale prototype reactor in regular operation, as well as during postulated accident scenarios. These reactor design analyses are generally performed using system codes and other associated simulation tools that require assessment, verification and validation using an appropriate experimental dataset. Experimental facilities used for reactor system safety analysis and system code assessments are categorized into integral effects test (IET) and separate effects test (SET) facilities. These IET and SET experiments and studies use geometrically scaled systems to reproduce the prototype system behavior at a reasonable cost, albeit with some scaling-related distortions. The design challenge of these model facilities is to identify and minimize scaling distortions while reproducing the most important operational phenomena in steady-state operation and in postulated accident scenarios. Lessons learned from previous experimental facilities, models, and correlations can support the development of new multipurpose, scaled, hybrid, integrated, and modular experimental facilities for advanced light water-cooled small modular reactors (SMRs). Successful operation of these facilities can significantly reduce upfront reactor development and demonstration costs and time to deployment.

Experiments on axial-flow-induced vibration of a free-clamped/clamped-free rod for light-water nuclear reactor applications

The flow-induced vibration of a cantilever rod confined in a pipe and exposed to an axial water flow has been experimentally investigated using a simple test piece designed to be informative for water-cooled nuclear reactor fuel rods. The Reynolds number was varied between 7.43 k and 82.5 k: a range that significantly extends the scope of existing data. Measurements were carried out using non-contact techniques: fast video imaging was used to track the rod motion, whilst particle image velocimetry was employed to resolve the flow field. The sources of excitation that sustain the rod vibration include turbulent buffeting, movement-induced excitation, and flow separation at the rod tip, whilst the observed dynamic responses of the rod comprise a fuzzy period-1 motion, a flutter-like large-amplitude oscillation, and buckling. Since both the rod vibration and the flow field are measured, the present data are particularly suited for numerical methodology development and validation.

Study on the costs and benefits of establishing a unified regulatory guidance for emergency preparedness of small modular reactors in China

Small modular reactors (SMRs) are rapidly developed worldwide in recent years for their potential extensive applications. China is very active in promoting SMR design and construction. However, the current emergency preparedness (EP) guidance in China is developed based on the characteristics of large water-cooled reactors, and they are only partially applicable to SMRs. In our previous work, we have fully demonstrated from a technical perspective that it is feasible for China to establish a unified EP framework for all SMR designs based on the dose and consequence of an accident spectrum. Therefore, this paper will comprehensively discuss the costs and benefits of establishing a unified EP guidance for all SMR designs in China. Uncertainty analysis is also conducted for variables with great uncertainty. Through the study in this paper, it can be concluded that developing a new unified EP regulatory guidance for all SMR designs in China is cost beneficial in most cases. In addition, it will also bring other benefits that are hard to be quantified, such as the consistency of supervision, the clarity of the EP plan, and the improved engagement of all stakeholders during SMR development. In a word, it would always be beneficial to establish a new unified EP regulatory guidance for all SMR designs in the long term.

Numerical analysis on subcooled boiling in PWR coolant channel based on a modified multi-scale interface model

A modified multi-scale interface model was employed with the Eulerian two-fluid framework to investigate the three-dimensional characteristics of subcooled boiling characteristics in a typical Pressurized Water Reactor (PWR) coolant channel with mixing grid. Based on the morphology identification by adopting an advanced interface topology map, the phase-average method and the interface-reconstruction method can be applied for dispersed bubble and sharp interface, respectively, to improve the calculated precision of interfacial transfers. Models were validated by using the dam break data for sharp interface topology and employing the flow boiling experimental data for the phase distribution in both radial and axial directions. The calculated results agreed well with the experimental data. Thermohydraulic characteristics in the coolant channel were obtained, including void fraction, phase velocity and flow streamline, based on which, the effects of mixing vane on the localized field were analyzed. The vapor volume fraction was decreased in downstream of spacer grid due to the mixing effect, while the lateral flow was enhanced. The vapor accumulation phenomenon was captured and the sharp interface between the liquid and the accumulated vapor was identified by the modified model. Based on this, the interfacial transfer characteristics of the sharp interface was considered in simulations.

Spray Cooling of Vertical Inline Horizontal Tubes

In water-cooled nuclear reactor safety, quench cooling is essential for minimizing core damage in the event of a Loss of Coolant Accident (LOCA). A hot, dry surface is quenched by bringing it into abrupt contact with a coolant, hence causing a fast fall in surface temperature. In this study, an air-atomized water spray cooling facility was built, and a series of tests were carried out to explore the quenching and rewetting of vertical inline horizontal tubes. During the cooling process, constant heat flux was provided on each tube. In the absence of a coolant, the steady heat flux represents the decay power of the channel components that reach the calandria tube surface. Temperature variations in the circumferential direction were measured over time during cooling. The effect of water pressure and air pressure on the rapid cooling of vertical inline horizontal tubes has been investigated. As air pressure rises, the quenching temperature of each tube decreases. A high-speed video camera was employed to investigate the flow behavior on the top tube surface as well as the spread of the re-wetting front around it. Experiments and visual analysis show that wetting front velocity rises with the water flow rate.

The Role of Determinism in the Prediction of Corrosion Damage

This paper explores the roles of empiricism and determinism in science and concludes that the intellectual exercise that we call “science” is best described as the transition from empiricism (i.e., observation) to determinism, which is the philosophy that the future can be predicted from the past based on the natural laws that are condensations of all previous scientific knowledge. This transition (i.e., “science”) is accomplished by formulating theories to explain the observations and models that are based on those theories to predict new phenomena. Thus, models are the computational arms of theories, and all models must possess a theoretical basis, but not all theories need to predict. The structure of a deterministic model is reviewed, and it is emphasized that all models must contain an input, a model engine, and an output, together with a feedback loop that permits the continual updating of the model parameters and a means of assessing predictions against new observations. This latter feature facilitates the application of the “scientific method” of cyclical prediction/assessment that continues until the model can no longer account for new observations. At that point, the model (and possibly the theory, too) has been “falsified” and must be discarded and a new theory/model constructed. In this regard, it is important to stress that no amount of successful prediction can prove a theory/model to be “correct”, because theories and models are merely the figments of our imagination as developed through imperfect senses and imperfect intellect and, hence, are invariably wrong at some level of detail. Contrariwise, a single failure of a model to predict an observation invalidates (“falsifies”) the theory/model. The impediment to model building is complexity and its impact on model building is discussed. Thus, we employ instruments such as microscopes and telescopes to extend our senses to examining smaller and larger objects, respectively, just as we now employ computers to extend our intellects as reflected in our computational prowess. The process of model building is illustrated with reference to the deterministic Coupled Environment Fracture Model (CEFM) that has proven to be highly successful in predicting crack growth rate in metals and alloys in contact with high-temperature aqueous environments of the type that exist in water-cooled nuclear power reactor primary coolant circuits.

Investigation of Stress Corrosion Cracking Resistance of Irradiated 12Cr Ferritic-Martensitic Stainless Steel in Supercritical Water Environment

The supercritical water-cooled reactors (SWCR) belong to Generation IV of reactors. These reactors have a number of advantages over currently operating WWERs and PWRs. These advantages include higher thermal efficiency, a more simplified unit design, and the possibility of incorporating it into a closed fuel cycle. It is therefore necessary to identify candidate materials for the SWCR and validate the safety and effectiveness of their use. 12Cr ferritic-martensitic (F/M) stainless steel is considered a candidate material for SWCR internals. Radiation embrittlement and corrosion cracking in the primary circuit coolant environment are the main mechanisms of F/M steels degradation during SWCR operation. Here, the stress corrosion cracking (SCC) in supercritical water at 390 and 550 °C of 12Cr F/M steel irradiated by neutrons to 12 dpa is investigated. Autoclave tests of specially designed disk specimens in supercritical water were performed. The tests were carried out under different constant load (CL), temperature 450 °C, and pressure in autoclave 25 MPa. The threshold stress, below which the SCC initiation of irradiated 12Cr F/M steel does not occur, was determined.

The influence of duct wall shape on mass redistribution in supercritical water-cooled reactor fuel assembly

Based on the fuel assembly of supercritical water-cooled reactor, a 19-rod bundle with triangular arrangement is established and different shapes of duct wall are used to adjust the hydraulic diameter of the external sub-channel. The Reynolds stress model (RSM) and enhanced wall treatment were employed to simulate the fluid flow and heat transfer phenomena of supercritical water and the mass redistribution in fuel assembly. The result shows that the hydraulic diameter difference between internal and external sub-channels and the physical properties of supercritical water are the main factors affecting the mass redistribution phenomenon. The mass redistribution can be effectively controlled by adjusting the hydraulic diameter of the external sub-channel. As the hydraulic diameter of the external sub-channel decreases, the fluid accumulation to the external sub-channel gradually weakens and then reverses to the internal sub-channel accumulation. Meantime, comparing with the primary model, the overall heat transfer effect of internal and external sub-channel can be balanced and kept at a high level by weakening the mass redistribution. Besides, the circumferential non-uniformity of the outermost fuel rod is affected by the local hydraulic diameter and the mass redistribution. Mass redistribution is the dominant factor and the local hydraulic diameter can only start to affect the circumferential wall temperature until the mass redistribution is weakened to a certain extent.

CFD prediction of heat transfer at supercritical pressure with rough walls: Parametric analyses and comparison with experimental data

In this paper, a low-Reynolds number turbulence model developed by the authors in past activities for rough walls is used for predicting heat transfer at supercritical pressure in the presence of different degrees of surface finishing. The model, able to reproduce the typical trends of friction factors from classical data reported by Nikuradse and summarised in the Moody diagram, is based on a simple-minded description of the effect of wall protrusions through the boundary layer on turbulence production.Though prior validation of the model only on the basis of friction factor data did not assure any basis for achieving accuracy in heat transfer prediction, a sensitivity analysis is firstly presented in order to characterise the obtained predictions at variable values of the roughness parameter, in particular concerning the possible suppression of deteriorated heat transfer by roughened surfaces. These analyses are extended to different fluids, making use of a fluid-to-fluid similarity theory recently proposed by the authors in order to establish similar boundary conditions and predicted phenomena.The results obtained by these analyses can be considered interesting, especially in view of the design of supercritical water-cooled nuclear reactors; however, an assessment against experimental data was obviously necessary. Experimental carbon dioxide data published in a very recent archival paper were thus addressed and were found useful in this regard. Considering these data allowed to extend the above analysis to provide confirmation of the promising features of the model in comparison with wall temperature values obtained with different boundary conditions. The model here described appears promising not only for its capability to predict experimentally measured effects, but also for the perspective to be used in the study of the behaviour of purposely roughened surfaces reducing the probability of occurrence of deteriorated heat transfer.

Numerical simulation of the fluid flow and heat transfer of the supercritical water in the 64-element Canadian SCWR fuel bundle

ABSTRACT Canada has participated in the Generation IV nuclear energy systems focusing on the Supercritical Water-cooled Reactor (SCWR) system. The 64-element fuel rods channel is used in the Canadian SCWR. There are few publicly available experimental studies for the flow in the fuel channels with multiple fuel rods. To date, the CFD simulation reported in open literature for rod bundle flow in the 64-element SCWR is relatively scarce. The cladding surface temperature is of key importance in assessing the safety of the reactor. Since the inhomogeneities in the bundle cross-section can present complex flow phenomena, a CFD study can provide substantial insight into the flow physics. In this work, the full-scale CFD simulation of the supercritical water flow in the 64-element rod bundle was performed. The results suggest the possibility of the presence of gap vortices in the flow subchannels. Higher streamwise velocities and normal Reynolds stresses always exist at the center subchannel regions. The circumferential cladding surface temperature distribution is extremely non-uniform and there is a large difference between the maximum cladding surface temperatures for different fuel rods.

Polycrystalline diamond and magnetron sputtered chromium as a double coating for accident-tolerant nuclear fuel tubes

Controlling the hot steam corrosion of nuclear fuel tubes is crucial for power plant efficiency and safety. We demonstrate an effective method of protecting ZIRLO ® (ZIRLO) fuel tubes against corrosion at accident temperatures in a water-cooled nuclear reactor environment. The tubes surfaces were coated by double layers consisting of the polycrystalline diamond (PCD) and magnetron-sputtered Cr. Coated and bare ZIRLO tubes were subjected to hot steam tests for 30 min at 900 °C followed by 40 min at 1000 °C. Surprisingly, Cr and PCD double layers ability to protect ZIRLO fuel tubes against hot steam corrosion strongly depends on the order of the layers in coatings. ZIRLO coated with Cr as the bottom layer and PCD as the top layer exhibited the best ability to slow down the oxidation process. This double layer coating reduced ZIRLO tube corrosion by 47% compared to unprotected ZIRLO. On the contrary, the double layer coating with PCD as bottom and Cr as the top layer reduced the ZIRLO tube corrosion by 20–25% compared to unprotected ZIRLO. To explain the specific roles of Cr and PCD coatings experimental and theoretical procedures were performed. Our results pointed to two basic parameters, significantly affecting the corrosion of coated ZIRLO tubes: 1. the amount of Cr carbides in the surface part of the coated ZIRLO fuel tubes and 2. adhesion of the coatings to the ZIRLO substrate.

Introducing Passive Nuclear Safety in Water-Cooled Reactors: Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres

Abstract The development of an accident tolerant nuclear fuel for water-cooled reactors would redefined the status of these reactors from traditional active safety to passive safety systems. As a possible solution toward enhancing the safety of light-water reactors (LWRs), loose-coated particles of enriched uranium dioxide (UO2) fuel with the ability to retain gaseous and metallic fission products in the case of a loss of cooling event can be introduced inside Silicon-Carbide cladding tubes of the fuel assembly (see Figs. 1(a) and 1(b)). These coated particles are treated as a bed from where heat is transferred to the cladding tube and the helium gas movement is due to natural convection. A slender geometrical model with tube-to-particle diameter ratio N = 2.503 and porosity ε = 0.546 mimicking the proposed nuclear fuel in the cladding was numerically simulated. This study is to investigate the heat transfer characteristics and flow distribution under buoyancy driven force expected in the cladding tube of the proposed nuclear fuel using a commercial code. Random packing of the particles is achieved by discrete element method (DEM) simulation with the aid of starccm+. The temperature contour and velocity vector plots obtained can be said to be good illustration of anticipated heat transfer and transport phenomenon to occur in the proposed fuel design. Simulated results for particle-to-fluid heat transfer coefficient, Nusselt number, and Rayleigh number which are of prime importance when analyzing natural convection heat transfer performance in fixed bed reactors were validated. Results from this work show close agreement with results obtained in established numerical and experimental works.

Tree-Based Ensemble Learning Models for Wall Temperature Predictions in Post-Critical Heat Flux Flow Regimes at Subcooled and Low-Quality Conditions

Abstract Accurately predicting post-critical heat flux (CHF) heat transfer is an important but challenging task in water-cooled reactor design and safety analysis. Although numerous heat transfer correlations have been developed to predict post-CHF heat transfer, these correlations are only applicable to relatively narrow ranges of flow conditions due to the complex physical nature of the post-CHF heat transfer regimes. In this paper, a large quantity of experimental data is collected and summarized from the literature for steady-state subcooled and low-quality film boiling regimes with water as the working fluid in vertical tubular test sections. A low-quality water film boiling (LWFB) database is consolidated with a total of 22,813 experimental data points, which cover a wide flow range of the system pressure from 0.1 to 9.0 MPa, mass flux from 25 to 2750 kg/m2 s, and inlet subcooling from 1 to 70 °C. Two machine learning (ML) models, based on random forest (RF) and gradient boosted decision tree (GBDT), are trained and validated to predict wall temperatures in post-CHF flow regimes. The trained ML models demonstrate significantly improved accuracies compared to conventional empirical correlations. To further evaluate the performance of these two ML models from a statistical perspective, three criteria are investigated and three metrics are calculated to quantitatively assess the accuracy of these two ML models. For the full LWFB database, the root-mean-square errors between the measured and predicted wall temperatures by the GBDT and RF models are 5.7% and 6.2%, respectively, confirming the accuracy of the two ML models.

Assessing techno-economic uncertainties in nuclear power-to-X processes: The case of nuclear hydrogen production via water electrolysis

Nuclear assisted low carbon hydrogen production by water electrolysis represents a potential application of nuclear cogeneration towards deep decarbonization of several fossil fuel-dependent industrial sectors. This work builds a probabilistic techno-commercial model of a water electrolysis plant coupled to an existing nuclear reactor for base load operations. The objective is to perform discounted cash flow (DCF) calculations for levelized nuclear hydrogen production cost under input parameter uncertainty. The probability distributions of inputs are used with the Monte Carlo-Latin Hypercube (MC-LH) sampling technique to generate 105 input scenarios and corresponding distribution of the levelized or life cycle hydrogen production cost instead of deterministic point values. Based on current techno-economic conditions, the levelized production costs of electrolytic hydrogen using electricity from large water-cooled nuclear reactors are determined to be US $ 12.205 ± 1.342, 8.384 ± 1.148 and 6.385 ± 1.051/kg H2 respectively at rated alkaline water electrolyser capacities of 1.25 MW(e), 2.5 MW(e) and 5 MW(e). The corresponding values for PEM water electrolysers are US $ 13.162 ± 1.356, 8.891 ± 1.141 and 6.663 ± 1.057/kg H2. The potential for flexible nuclear reactor operation and management of power demand uncertainties through nuclear hydrogen cogeneration is also examined through a case study.

Review of Euratom projects on design, safety assessment, R&D and licensing for ESNII/Gen-IV reactor systems

Five Euratom projects launched since 2017 in support of the development of ESNII/Generation-IV reactor systems are briefly presented in the paper in terms of key objectives, results, and recommendations for the future. These projects focus on various aspects of the following ESNII/Generation-IV systems: Sodium Fast Reactor, Gas Cooled Fast Reactor, Supercritical Water Cooled Reactor, and Molten Salt Fast Reactor. The paper does not consider EU projects focused on the Gen-IV reactor technologies based on the use of heavy metals as a coolant because these projects are reviewed in a different paper.