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• Use two advanced algorithms to optimize and improve the network. • Use the parameter fine-tuning strategy to make the Transfer learning method more suitable for the target field. • The adopted Data and information visualization library can intuitively show the classification effect of the model. Deep convolutional neural networks (DCNN) are widely applied in the realm of deep learning. This paper presents a novel approach that combines transfer learning techniques with a hybrid domain attention mechanism module to enhance and refine the DCNN architecture, consequently boosting its performance. The focus of this study is the application of the improved DCNN model to fault diagnosis within small modular reactor. We hope to improve the fault monitoring and diagnosis capabilities of small modular reactors through algorithm improvements, to enhance their safety. Comparative results demonstrate that the improved model surpasses other deep learning models in terms of convergence rate, recognition accuracy, and model size, evidencing robust generalisation capabilities. [ABSTRACT FROM AUTHOR]

The current study develops a hybridized small modular nuclear reactor and solar-based system designed specifically for sustainable communities in metropolitan areas to meet their power, heat, clean fuel, fresh water and food requirements. Both floating-type and bifacial-type photovoltaic arrays are integrated with a high-temperature gas-cooled small modular reactor. In the integrated system, a Rankine cycle subsystem for power generation, a solid oxide electrolyzer for hydrogen production, a pressure swing adsorption unit with ammonia reactor for nitrogen and then ammonia production, a multi-effect desalination unit for freshwater generation, and heat recovery units in order to provide heating to residential area, food drying facility, greenhouse, and fish farm facilities. For both floating and bifacial PV options, the 120 MWp PV modules are considered and integrated with two small modular reactor units at 250 MWth capacities each, 500 MWth in total. A thermodynamic analysis, which is based on energy and exergy approaches, is carried out. The time-dependent analyses are carried out for floating-type PV and bifacial-type PV options for the cities of Istanbul in Turkey; Barcelona in Spain; Los Angeles (LA) in the United States; Tokyo in Japan; and Toronto in Canada by using their hourly meteorological data, source and load capacities in a typical meteorological year. Among ten cases, between 112835 and 146180 tonnes of ammonia are generated in a typical year while meeting the heating and power requirements of sustainable communities in metropolitan areas. Among ten cases, the average overall energy and exergy efficiencies are found as 41.04% and 46.88%, respectively. The hybridization of such nuclear and solar systems achieves unique advantages by preventing intermittencies compared to the renewable-alone systems, reducing carbon emissions compared to fossil-based systems, and lowering the levelized costs of energy compared to the nuclear energy-alone systems. • Solar and small modular reactor-based system is developed for multigeneration. • Floating and bifacial PV options are investigated within the integrated system. • The proposed system is designed in a way to fit urban environment. • A time-dependent analysis is carried out with energy and exergy approaches. • The average overall energy and exergy efficiencies are found as 41.04 % and 46.88 %. [ABSTRACT FROM AUTHOR]

• Pyrex and CIMBA-loaded cores are designed and compared for BANDI-60 reactor. • Both cores use soluble boron-free operation for reduced size and enhanced safety. • Core designs are simulated using McCARD and neutronics parameters are compared. • Both cores demonstrate an approximate cycle of 4.9 effective full-power years. • Both control element assemblies can guarantee shutdown margins. The BANDI-60 Soluble Boron-Free Small Modular Reactor (SMR) developed by KEPCO E&C is a Pressurized Water Reactor producing 200 MW of thermal power. This study proposes core designs for the BANDI-60 using two types of burnable absorbers (BAs) and compares their characteristics. One core uses the discrete Pyrex BA, and the other uses a new BA design named "cylindrically inserted and mechanically separated burnable absorber" (CIMBA). Pyrex is the borosilicate glass consisting of B 2 O 3 -SiO 2 , which suppresses excess reactivity to achieve a long cycle length. CIMBA is a wire-type gadolinia (Gd 2 O 3) BA inserted into annular fuel pellets, which utilizes a spatial self-shielding effect on the fuel material and has a greater fuel inventory than discrete BAs. The core designs are performed using a McCARD code, and the results showed both cores can operate over an approximate cycle length of 4.9 years and possess a reactivity control capability ensuring a safe shutdown. [ABSTRACT FROM AUTHOR]

• Revised boiling heat transfer model of the secondary loop of MCOTSG. • Using other correlation to revise the Chen correlation. The mini-channels casing-pipes once-through steam generator(MCOTSG) is used for a certain type of small modular reactor. The boiling heat transfer model is crucial to the accuracy of simulating the heat transfer process on the secondary loop of MCOTSG. In this paper, the heat transfer tubes of MCOTSG are taken as the research object. The thermal–hydraulic model of primary and secondary loops is established based on the RELAP5 program. Then the different nucleate boiling heat transfer models are calibrated and analyzed based on the experimental results of the steady state under low flow rate conditions. On this basis, the secondary loop heat transfer characteristics and the overall operating characteristics of OTSG under the transient state of feedwater flow rate reduction are simulated. The results of the steady state and transient state analysis show that the revised model that the Bertsch correlation is used to correct the Chen correlation performs the best in terms of computational stability and computational accuracy. The results of the study can provide a basis for the safety analysis and optimal design of MCOTSG and small modular reactor under low flow rate conditions. [ABSTRACT FROM AUTHOR]

The transition toward a low-carbon energy system and the increasing penetration of variable renewable energy (VRE) sources translate into a pressing need for dispatchable and low-carbon power sources. Nuclear hybrid energy systems (NHES) exploit the synergies between nuclear power and other energy sources together with energy storage devices and a variety of electric and non-electric applications. The expected benefits range from a high flexibility being able to supporting an increasing penetration of the VRE while complying with the grid demand and constraints to an increased profitability brought by the production of commodities beyond electricity (e.g., hydrogen, heat, etc.). A dedicated framework must be developed to evaluate different NHES configurations, particularly with regard to the complex interconnections among the tightly coupled components. In this work, illustrative examples of NHES components were selected and modeled with the object-oriented modeling language Modelica and implemented in the Dymola simulation environment. The technologies considered in this study are a Small Modular Reactor (SMR) based on pressurized water technology, a thermal energy storage (TES) system, and an alkaline electrolyzer for hydrogen production. The dynamic models are then collected in a new Modelica library and assembled into a variety of NHES topologies using a plug-and-play approach. The time-dependent behavior of the NHES layout can be simulated under different operational contexts, enabling the monitoring of key process variables, supporting system design, exploring alternative control strategies, and analyzing different scenarios. The NHESs are investigated in two exemplary scenarios – one representing typical load conditions and the other featuring high VRE penetration – in order to demonstrate the viability of the proposed approach as an initial effort toward the development of a holistic framework for analyzing NHES. The dynamic models effectively met the analysis requirements, for instance, by tracking the production of commodities throughout each operational transient, which is an essential result for evaluating the performance of NHES. In this regard, efficiency is adopted as the figure of merit to compare the different NHES architectures, with simulation results indicating significant overall efficiency improvements in NHES incorporating TES and using nuclear heat to drive non-electric applications. • Dynamic modeling of five nuclear hybrid energy system architectures in Modelica. • Integration of a SMR, a thermal energy storage system, and an alkaline electrolyzer. • Methodology overview to assess hybrid systems' performances in different scenarios. • Explores option of meeting load demand while operating SMR at nominal power. [ABSTRACT FROM AUTHOR]

• A boron-free SMPWR core design with helical-cruciform metallic fuel is proposed. • The reactivity control optimizations of burnable absorbers and control rods are analyzed with genetic algorithm. • The NETH-HCF175M core can achieve a cycle length of about 1360 EFPD. Small modular pressurized water reactors have the advantages of small size, modularity, inherent and passive safety, and flexible power generation for a wider range of users and applications. Helical-cruciform metallic fuel has the potential to be applied to small modular pressurized water reactors capable of its axial self-spacing plane, high power density, increasing heat transfer area, and extending the cycle length. A novel boron-free small modular pressurized water reactor NETH-HCF175M design using helical-cruciform metallic fuel is proposed in this paper. The reactivity of the core is completely controlled by burnable absorbers and control rods. Two different geometric structures of separate pins and integral pins are researched. Results indicate that Gd 2 O 3 is a suitable neutron absorbing candidate, while Er shows the best reactivity control effect in the integral pin structure. The separate pin structure with Gd 2 O 3 is selected as the burnable absorber design and the core depletion result shows that the core can achieve a cycle length of about 1360 effective full power days. Control rods are divided into 60 compensating rods and 60 shutdown rods for excess reactivity controlling and shutting the core. The genetic algorithm is applied to the optimization work of layout and concentration design for burnable absorbers and layout of compensating rods. Based on the optimized burnable absorbers and compensating rod designs, excess reactivity can be controlled and the core can achieve criticality at beginning of cycle. The core can be transitioned from hot full power to cold zero power with shutdown rods at BOC, which obtains a shutdown margin more than 3000 pcm. Finally, the physical characteristics of core are analyzed, including depletion simulation, pin power distribution, and reactivity equilibrium. This paper provides an overview and information practical for the reactivity control design of novel SMPWR with high-enrichment and geometric complex helical-cruciform metallic fuel. [ABSTRACT FROM AUTHOR]

• The natural convection at high Rayleigh number (Ra ≈ 1015) is simulated by LES method. • The calculated correlation Nu ∝ Ra 0.33 at Ra ≈ 1015 can reproduce the classical 1/3 law. • The turbulent regime transition from the buoyancy dominated to the shear stress dominated at high Rayleigh number may confirm previous theoretical prediction. • The turbulent vortex structure at high Rayleigh number is dominated by streak-like flow and the streak scale grows with the increase of Rayleigh number. The passive safety concept of Small Modular Reactors (SMR) is based on the transfer of residual heat from the reactor to a water pool. Since the height of the reactor vessel reaches approximately 15 m , this leads to a strong heat exchange between the wall and the water by the way of natural convection. The maximum Rayleigh number (Ra) reaches 10 14 or 10 15. Reliable heat transfer correlations exist only up to Ra ≈ 10 12 , still along with uncertainties in the extrapolation to higher Rayleigh number values. To improve the understanding of natural convection at high Rayleigh number values, the turbulent natural convection boundary layer (TNCBL) along a 15 m high, side-heated vertical wall in a water tank is simulated by employing the Large Eddy Simulation method with the CEA in-house code Triocfd (Angeli et al., 2015). The numerical results in terms of heat transfer correlation, mean temperature and mean velocity, second moment variables show a reasonable agreement with available experimental data in the range of Ra < 10 12. It is observed in the simulation that the powers of the heat transfer correlations in the laminar (Nu y = 0.56 Gr y ∙ P r 0.25) and the turbulent regimes (Nu y = 0.114 Gr y ∙ P r 1 / 3) correspond well to the classical regimes at Ra < 2 × 10 12. A slight deviation from the classical exponent (1/3) is found at higher Rayleigh number (2 × 10 12 < Ra < 1.8 × 10 15), namely Nu y = 0.114 (Gr y ∙ P r) 0.33 . Through the analysis of turbulent kinetic energy budget, it is found that turbulent contribution from shear stress increases continuously with the increase of Rayleigh number. However, the turbulence contribution from buoyancy keeps increasing until Ra ≈ 6.6 × 10 13 , then decreases gradually at higher Rayleigh number. The behavior of the maximum buoyant force shifts slowly from the inner boundary layer to the outer boundary layer may indicate the potential transition of the buoyancy-dominated regime to the shear stress dominated regime, thus confirming previous theoretical prediction (Wells Andrew and Worster, 2008). The increasing scale of the streak-like turbulent structures with the increase of Rayleigh number are clearly demonstrated through visualization process. [ABSTRACT FROM AUTHOR]

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead–bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Most of the small modular reactors (SMRs) under development worldwide present the same components: an integral reactor vessel with a low-positioned core as the heat source and a high-positioned steam generator as the heat sink. Moreover, some SMRs are being designed to be driven by natural circulation during normal power generation. This work focused on such designs and on their performance, considering the changes generated by the geometric and hydraulic parameters of the system. Numerical simulations using mass, momentum, and energy equations that considered buoyancy forces were performed to determine the effects of various geometric and hydraulic parameters, such as diameters and flow resistances, on the reactor’s performance. It was found that nonuniform diameters promote velocity changes that affect the natural circulation flow rate. Moreover, the reactor’s temperature distribution depends on the steam generator tube pitch. Therefore, the hydraulic diameters of the reactor’s coolant passages should be maintained as uniform as possible to obtain a more uniform temperature distribution and a larger mass flow rate in SMRs.

Small Modular Reactors (SMRs) have been recognized to have potential in decarbonizing district heating, which is currently an urgent sustainability challenge in many European countries. In this paper, the financial impacts of replacing peat and biomass-powered Combined Heat and Power (CHP) plants with heat-only reactors of 24–200 MW power range and maximum output temperatures of 120 °C are studied. A district heating system of a medium-sized Nordic city in Eastern Finland is modeled with EnergyPRO computer software (EMD International A/S, Aalborg, Denmark), which is used to optimize plant units’ production for cost effectiveness. A future scenario is used to predict electricity prices, expenditures from CO2 emission allowances, and fuel prices for the studied case. Results show that the low operating expenditures of CO2 free heat-only reactors would compensate for the revenue losses from electricity sales and that a small number of micro reactors, with power output in the tens of megawatts range each, would be optimal for the studied case. Since investment cost estimates for SMRs still bear significant uncertainties, the subject should be followed in further studies, as heat-only SMRs could provide a profitable alternative for current CHP production in the future.

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

Summary: A core design of small modular liquid‐metal fast reactor (SMLFR) cooled by lead‐bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off‐grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low‐enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S‐CO2) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the core with low‐enriched uranium (LEU) performs the function of core ignition. The center region plays the role of a breeding blanket to increase the core lifetime for long cycle operation. The core working fluid inlet and outlet temperatures are 300°C and 422°C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi‐static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal‐hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the conceptual core is favorable in terms of self‐controllability, which is the first step towards inherent safety. A core design of small modular liquid‐metal fast reactor cooled by lead‐bismuth eutectic is developed with a thermal power of 37.5MW, a cycle length up to 30 years and small size (1m active height for 1.45m active radius) so that it can be transported in a fuel transport cask or used as a power source in an icebreaker ship. The significant neutronics, thermal hydraulics and safety parameters are analyzed, showing that the innovative core design can satisfy advanced requirements of GEN‐IV reactors such as simplicity in design, reliability for long operation time and prolifration resistance. [ABSTRACT FROM AUTHOR]

As one of the future directions of nuclear energy development, small modular reactor (SMR) designs meet the demands of safety, sustainability, and efficiency by eliminating circulating pumps and using natural circulation-driven flows to transfer fission energy to power. However, natural circulation-driven flows could be affected by two-phase-flow instability that may occur during accidental scenarios of pressurized water reactor (PWR)-type SMRs due to relatively small driving force. In view of the influence of two-phase-flow instability during accident transients for a PWR-type SMR, experiments are performed in a well-scaled test facility to investigate potential thermal-hydraulic flow instabilities during blowdown events. The test facility has a height of 3.44 m, and the operating pressure limit is 1.0 MPa. The scaling analyses ensure that the scaled phenomena, i.e., depressurization of the reactor pressure vessel (RPV) and emergency core cooling system valve actuation, could be accurately simulated in the test facility. Important thermal-hydraulic parameters including RPV pressure, containment pressure, local void fraction and temperature, pressure drop, and natural circulation flow rate are measured and analyzed during the blowdown events. Test results show that throughout the experiment the liquid level is always maintained above the heated core and the RPV pressure decreases. Oscillations of the natural circulation flow rate, water level, and pressure drop are observed during blowdown transients. Specific reasons for and mechanisms of the observed instability phenomena are discussed. [ABSTRACT FROM AUTHOR]

This work presents a neutronic assessment to convert a Small Modular Reactor (SMR) with uranium core to the thorium mixed oxide core with minimum possible changes in the geometry and main parameters of SMR core. This option is due to most of SMR are designed to be strongly poisoned in the beginning of cycle and to have a long cycle. Thorium can be used as an absorber in the beginning of the cycle and also be used as a fertile material during the cycle, it seems to be a good option to use (Th/U)O 2 as SMR’s fuel. The main neutronic objectives of this study is achieving longer cycle length for SMR by using the minimum possible amount of burnable poison and soluble boron in comparison with reference core. The Korean SMART reactor as a certified design SMR has been chosen as the reference core. The calculations have been performed by MCNP code for homogeneous and heterogeneous seed and blanket concept fuel assemblies. The results obtained show that the heterogeneous fuel assembly is the one which gives longer cycle length and used lower amount of burnable poison and soluble boron, and also consumes almost the same amount of 235 U. [ABSTRACT FROM AUTHOR]

In advanced small modular reactors, a natural circulation system with liquid metal is preferred, as it offers enhanced safety features, simplicity of reactor design, and is economical with respect to construction. A natural circulation system also provides potential non-proliferation benefits to the international community. To test lead-bismuth eutectic as a liquid metal coolant, we have studied its natural circulation characteristics. We present the results of experiments on natural circulation conducted in a full-height scale test loop for varying core power, heat exchanger coolant oil temperature, and bubble injection rate. The experiments on the liquid metal were conducted under core power source conditions ranging from 2.5 kW to 15 kW. The aim of this experiment was to study natural circulation under various conditions. With our demonstration facility-based technology, we plan to verify small modular reactor cooling system safety and the technology of the bubble injection system that has been developed for liquid metal flow devices. [ABSTRACT FROM AUTHOR]

The critical heat flux (CHF) is one of the main thermal-hydraulics safety limits in water-cooled reactors. At CHF, a film of vapor is formed on the heated wall and the wall temperature can sharply increase in a short period of time, which may lead to damage of the heater surface. In spite of many studies, the CHF mechanisms are still not well understood due to the complexity of this two-phase phenomenon and its dependence on local thermal-hydraulic conditions. Two-phase boiling experiments have been conducted to observe CHF and post-CHF behavior in prototypical conditions for small modular reactors (mass flux: 560–1570 kg/m 2 .s, pressure: 7.6–16.4 MPa, inlet subcooling: 160–440 kJ/kg) in a 2 × 2 square rod bundle geometry with a chopped-cosine power profile at the University of Wisconsin-Madison. The fuel rods temperature history during and following CHF were used in the solution of the two-dimension inverse heat transfer problem to estimate the wall temperatures and wall heat fluxes. This analysis showed different transient boiling curves for different flow regimes that differ qualitatively and quantitatively from the typical boiling curve considered in steady-state two-phase heat transfer analysis. The experiments suggest that the inverse heat transfer analysis approach can be used to estimate the post-CHF heat transfer and to better understand the CHF mechanisms at high pressure typical of water-cooled reactors. [ABSTRACT FROM AUTHOR]

Several cesium and strontium bioaccumulation models are used widely in national and international guidance for ecological and human health risk assessments for radiocesium (134Cs and 137Cs) and radiostrontium (90Sr), but have not been used to make predictions of radiological risk from nuclear accidents under variable environmental conditions on broad geographical scales. In this paper, we first present models for predicting the bioaccumulation of cesium and strontium by aquatic biota based on ambient concentrations of dissolved potassium and calcium, respectively, and then test these models using independent data from aquatic ecosystems at Canadian nuclear sites. Secondly, models yielding the best predictions across a wide range of input parameters were selected to estimate bioaccumulation factors (BAFs) for cesium and strontium in aquatic ecosystems across Canada, using trophic level of organisms and dissolved potassium for cesium and calcium concentrations for strontium as predictor variables, and presented as contour maps of radiological risk. The models show that risk from bioaccumulation of cesium and strontium can vary by several orders of magnitude depending on site-specific environmental conditions and trophic ecology. Overall, our results suggest that single-parameter approaches taken by regulatory standards may either over- or under-predict radiological risk at many locations, and are thus not readily suitable to inform siting decisions for new nuclear developments. [ABSTRACT FROM AUTHOR]

In support of more efficient utilization of solar and nuclear energy in power generation, the present work proposes a conceptual design of a hybrid nuclear-solar power system (HNSPS) for on-demand power supply, based on a parallel thermal integration of small modular reactors with commercialized molten-salt concentrating solar power tower plants. A cost-effective thermal energy storage system is employed as a thermal coupling and a heat dispatcher to coordinate both nuclear heat and solar heat with power demands. The overall configuration and operation strategy of the hybrid system are elaborated. A numerical model of the hybrid system is developed based on the circulation of molten-salt. Using the model, a 7-day performance analysis of the hybrid system with different considerations are conducted. Effects of key design parameters on system performances are investigated by a parametric study. A customized evaluation parameter, named as “design score”, is defined to assess the suitability of the designed configurations. The performance analysis shows that the integration of nuclear power with concentrating solar power allows the system to be less dependent on heat dispatch to achieve required power regulations, comparing to a standalone CSP plant with an equivalent power capacity. The parametric study indicates that a higher heat generation utilization of the hybrid system requires a larger TES and a smaller heliostats field, while a higher power supply efficiency of the hybrid system is accompanied with a larger TES, a larger heliostats field, and a higher nuclear power capacity proportion. The optimum design configuration of a 200 MWe HNSPS is estimated to have a nuclear power capacity proportion of 50%, a theoretical thermal energy storage duration of 14.8 h, and a heliostats field with a solar multiple of 1.27. This configuration can completely satisfy the power demand without any heat generation curtailment during the 7-day operation. The obtained results allow the HNSPS to be considered as a promising non-carbon-emitting power generation system for on-demand power supply. [ABSTRACT FROM AUTHOR]

In this paper, a market guided design approach is exercised to identify the optimal reactor form in terms of safety, cost, fuel availability, and profitability. Various Gen IV reactor concepts and existing reactor examples have been examined and potential non-power applications of nuclear energy is investigated. The conclusion drawn upon the state-of-the-art reactor designs is that thorium-base molten salt small modular reactor is the most promising technical path to carry nuclear energy into the new era of meeting the market demand while maintaining utmost safety at an affordable capital cost. [ABSTRACT FROM AUTHOR]

Nuclear power plants should be capable of contributing safely and flexibly to the electrical grid stability. This research deliberates the feasibility of a new innovative concept of a passive and autonomous frequency control operation (PAFO) in soluble-boron-free (SBF) pressurized water reactors (PWRs). The new concept eliminates any required active control in the reactor core for attaining primary and secondary frequency control operational modes in PWRs. In particular, control rods and soluble-boron are not utilized in a PAFO. Meanwhile, the reactor power maneuvering is achieved entirely by the passive reactor feedback to the coolant temperature change. This approach is motivated by the strongly negative coolant temperature coefficient available in an SBF coolant system. In order to ascertain judiciously the practical viability of the proposed concept, numerical simulations of passive primary and secondary frequency operations are performed using a lumped PWR system model solved by an in-house FORTRAN-95 computer code. The various simulation parameters are obtained from 3D Monte Carlo simulations of an SBF 450 MWt PWR by using Serpent-2 code with ENDF/B-7.1 data library. The numerical simulations are carried out at the beginning, middle and end of the fuel cycle to emphasize the effect of fuel burnup on the proposed PAFO. Moreover, sensitivity analyses on xenon worth and nominal fuel temperature are performed to deliberate their impacts in tailoring the reactor performance. Because the reactor power matches well with the power demand with a small variation in coolant temperature during the frequency control, we conclude that there is promising potential for the proposed PAFO concept. [ABSTRACT FROM AUTHOR]

Part of the mission for the Nuclear Advanced Manufacturing Research Centre (Nuclear AMRC) is to develop, mature and deploy advanced manufacturing technologies in the nuclear manufacturing industry. Electron beam welding (EBW) is widely considered to be one of the game-changing technologies which could reduce pressure vessel fabrication lead time, unit cost and manufacturing risk, while maintaining or improving quality & component integrity. This paper presents development work for the use of EBW as a joining technique for nuclear pressure vessel (NPV) steel walls of up to 100 mm thickness. SA508 Grade 3 as well as the higher strength type SA508 4N with increased hardenability were investigated. It has been demonstrated that EBW has the capability of joining these sections in a single pass, with elementary microstructural and mechanical examinations underpinning the basic performance support for such capability. Two-thirds scale components, such as those typical of a Small Modular Reactor (SMR) have demonstrated this processing duty-cycle further with the joint being completed in 26 minutes. This work will strengthen the viability of EBW applied to primary circuit safety-critical components, and will support a future ASME III code case. [ABSTRACT FROM AUTHOR]

The increased penetration of intermittent renewable energy technologies such as wind and solar power can strain electric grids, forcing carbon-based and nuclear sources of energy to operate in a load-follow mode. For nuclear reactors, load-follow operation can be undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various methods of thermal energy storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid variability caused by daily load demand changes and renewable intermittency. Two TES techniques are investigated as candidate thermal reservoirs to be used in conjunction with a small modular reactor (SMR): a two-tank sensible heat storage system and a stratified chilled-water storage system. The goal when coupling the two systems to the SMR is to match turbine output and demand and bypass steam to the TES systems to maintain reactor power at approximately 100%. Simulations of integral pressurized water reactor dynamics are run in a high-fidelity FORTRAN model developed at North Carolina State University. Both TES systems are developed as callable FORTRAN subroutines to model the time-varying behavior associated with different configurations of these systems when connected to the SMR simulator. Simulation results reveal the sensible heat storage system is capable of meeting turbine demand and maintaining reactor power constant while providing enough steam to power four absorption chillers for chilled-water production and storage. The stored chilled water is used to supplement cooling loads of an adjacent facility. [ABSTRACT FROM AUTHOR]

Experimental study on natural circulation flow instabilities is of great importance for the safety analysis in a PWR-type SMR, especially for accident scenarios such as loss of coolant accident (LOCA) and loss of heat sink accident (LOHS). In this study, an experimental natural circulation facility was built by scaling down from a typical PWR-type SMR. The scaling ratios were derived from non-dimensional field and constitutive equations of the drift flux model. The test facility has a height of 3.44 m with an operating pressure limit of 1.0 MPa. Two kinds of tests, the blowdown test and cold-blowdown test were performed. The blowdown test was designed to simulate the low pressure phase (<1 MPa) of a LOCA event and evaluate the performance of the emergency core cooling system (ECCS). The test results indicate that the ECCS can keep the RPV water level above the core, and the RPV pressure keeps decreasing during the entire accident transients. The cold-blowdown test was designed to simulate the whole transients of the system blowdown at low pressure conditions. The oscillations of the pressure drop, natural circulation rate, void fraction and inlet temperature were observed and analyzed. The cold-blowdown test results were compared with the RELAP5 simulation. Although some discrepancies existed at the initial blowdown phase, the code calculation agreed well with the experiment data. [ABSTRACT FROM AUTHOR]

Summary: A 20 MWth, 540 EFPD once through fuel cycle small modular molten salt reactor with solid fuel is proposed by Massachusetts Institute of Technology for off‐grid applications. In this paper, various thermal‐hydraulic analysis methods including computational fluid dynamics, Reactor Excursion Leak Analysis Program (RELAP5), and DAKOTA are adopted step‐by‐step for the reactor design based on the neutronic analysis results. First, 1/12th full core thermal hydraulic analysis is performed by using STAR CCM+ with most conservative considerations. Second, the transient safety behaviors of reactor system with risky assumptions are conducted by using REALP5. Finally, due to the unknown factors affecting reactor thermal‐hydraulic characteristics, the uncertainty quantification and sensitivity analysis for the designed reactor is performed with DAKOTA code coupled with RELAP5. Numerical results show that a more uniform temperature distribution with reduced peak temperatures of fuel and coolant across the reactor core has been achieved. Enough safety margin is maintained even under most severe transient accident. The uncertainties in the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. All above results preliminarily indicate the feasibility of the current small modular molten salt reactor design and provide the further optimization direction from reactor thermal‐hydraulic prospective. [ABSTRACT FROM AUTHOR]

• A Monte Carlo simulation has been conducted for various dispersed fuel compositions of reactor KLT-40S. • Use of (Th232 + U233)O 2 fuel composition increases fuel lifetime by 450 EFPD and fuel burnup by 45 MW∙d/kg HM. • (Th232 + U233)O 2 fuel composition at 7.2 mm diameter results in a 45% fuel lifetime increase and a 5% fuel burnup decrease. • (Th232 + U233)O 2 composition requires less volume of moderator at fuel rod diameters of 7.2 mm and 8.2 mm. • All fuel compositions, except for (Th232 + U233)O 2 , show optimal fuel cycle parameter values at project fuel rod diameter. This paper demonstrates the results of the Monte Carlo simulation conducted with MCU-PTR software for a small modular pressurized water reactor KLT-40S. Four dispersed fuel compositions, (U238 + U235)O 2 , (U238 + Pu239)O 2 , (Th232 + U235)O 2 , and (Th232 + U233)O 2 , were examined for fuel lifetime and fuel burnup calculations. The transition to the (Th232 + U233)O 2 fuel composition extends fuel lifetime by 50% and fuel burnup by 44% at design value of the fuel rod diameter. If the fuel rod diameter for (Th232 + U233)O 2 fuel composition is increased to 7.2 mm, fuel lifetime is extended by 105%, and fuel burnup is extended by 38% compared to the design values. [ABSTRACT FROM AUTHOR]

• Compared to the typical pressurized water reactor, an advanced reactor type is coupled with seawater desalination. • Four options with MED and RO desalination technologies are evaluated and compared with parametric studies. • The performance of the energy utilization, production flexibility of water and electricity and water quality are considered as criteria. • The RO option has the best performance of energy utilization and a higher upper limit of water production. • The MED medium-pressure option could be an alternative when the 3 criteria are all required. To respond to the increasing water demand worldwide, especially in coastal cities and islands, nuclear desalination becomes a potential solution. SmTMSR-400 is a small modular molten salt reactor, which can be easily transported and has inherent safety. This paper analyzes the options of coupling SmTMSR-400 with desalination with parametric studies. SmTMSR-400 is integrated with multi-effect distillation (MED) and reverse osmosis (RO), respectively. In the MED options, the high-pressure, medium-pressure, or low-pressure steam from the steam Rankine power cycle can be used to generate the process steam for desalination. After evaluating and comparing the four options, it is found that: a) the RO option has the best performance of energy utilization and a higher upper limit of water production; b) the RO option is not suitable for water sources containing a large amount of organic matter, and the quality of the water produced is lower than that of the thermal desalination options; c) the low-pressure steam option has the best performance of energy utilization among the MED options, but its water production and power generation cannot be adjusted according to the local demand; d) when the performance of energy utilization, production flexibility, and water quality are all required, the medium-pressure steam option could be an alternative. [ABSTRACT FROM AUTHOR]

The relevance of nuclear power plant flexible power operation (FPO) is rising due to increased penetration of variable renewables. Small Modular Reactors (SMRs) such as NuScale are envisaged to have multiple units on one site. This provides opportunities for coordinating FPO between units, but also introduces challenges as common services (e.g., refueling equipment) must be shared between units and fuel loadings may be standardized. It is therefore important to quantify whether FPO with SMRs is economically beneficial. This paper quantifies the economics of FPO on different timescales to analyze refueling outages of such multi-unit SMRs. A 24-h price-taker profit maximization of SMR operation is first solved considering revenues from wholesale power and ancillary service (AS) markets. The results are then used to find the most profitable operation strategy over several years for multiple units at the site accounting for the physics and materials limits on NPP operation, including refueling outages. Results show a small but appreciable participation of nuclear into AS, contributing 4% of revenues. Due to the longer fuel cycle, over a decade of operation the units' refueling outages drifted by 3 months, ultimately leading to refueling during summer. • We develop a model for multi-unit small modular nuclear reactor operation. • Mixed integer linear programming is used to optimize for revenue. • Short and long timescale models are used in combination. • It is found that flexible power operation is economically beneficial. • Outage scheduling drifts over time due to longer fuel cycle. [ABSTRACT FROM AUTHOR]

Small Modular Reactors are gaining significant interest for their reduced footprint, lower power output, modularity, and innovative features. The licensing of SMRs is key to their successful deployment. However, the literature on this subject area is limited and often fragmented among other characteristics of the SMRs, thus failing to address the licensing aspects distinctly. The paper employs a systematic literature review to identify the potential nuclear licensing barriers and challenges that can influence the deployment of SMR and to provide an overview of their implications. The authors differentiate between licensing barriers and challenges as follows. The licensing barriers are likely to affect the deployment of SMRs for over a decade and necessitate the collaboration of multiple organisations. The licensing challenges can be resolved within ten years and can be led by a single organisation to deliver the solution. The licensing barriers are: (1) existing legal and regulatory framework; (2) prescriptive regulatory framework; (3) novelty in the technology; (4) regulatory fragmentation; and (5) absence of in-factory certification. The licensing challenges are: (1) fees charged by regulators; (2) regulatory capability gaps; and (3) lengthy licensing duration. The identified barriers and challenges have implications on the project timeline and cost, consequently affecting the overall economics of the SMR. [ABSTRACT FROM AUTHOR]

Advancements in computer performance and Artificial Intelligence, especially Deep Neural Networks and reinforcement learning, have caught the attention of many researchers. Specifically, the potential for achieving autonomy (a higher cognitive concept than legacy automation systems) has garnered significant interest. Thus, many autonomous systems like self-driving cars, are in the research/ development phase. Despite its numerous advantages, the adoption of autonomous control for Nuclear Power Plants is deficient mainly due to stringent safety concerns. Especially, autonomous control for Small Modular Reactors is a desirous capability to accrue its full potential. This study proposes an AI-based model for autonomous control of IP-200 NPP, a Small Modular Reactor (SMR) under development. The model adopts a modular architecture for performing different functions of monitoring/ diagnosis, strategy formulation and assessment. Despite following knowledge-based rules, the network decisions are optimized through supervised learning and fine-tuned through reinforcement learning. The results demonstrate that the proposed model can handle plant emergencies while ensuring plant safety at optimal performance. [ABSTRACT FROM AUTHOR]

• Thermophysical analysis of KLT-40S fuel assembly has been performed. • Fuel burnup values have been obtained for different dispersive fuel compositions and outer fuel rod diameters. • Kutateladze's method was applied to include effect of subcooled boiling and refine the cladding wall temperature. • Altering fuel composition to Th-U233 slightly decreases maximal temperatures of both fuel and cladding. • Increasing the outer fuel rod diameter up to 7.2 mm improves the thermophysical safety of the reactor core. This paper contains the results of KLT-40S reactor unit thermophysical studies and describes its main design features. Such quantities as fuel burnup, Reynolds and Prandtl numbers, coolant flow velocity, temperature distributions for fuel and fuel rod cladding, and departure from nucleate boiling ratio up to boiling crisis were calculated for different fuel rod diameters and fuel compositions. It is shown that changing the dispersed fuel composition from (U238 + U235)O 2 to (Th232 + U233)O 2 does not alter the maximal temperatures of fuel and fuel rod cladding significantly. An increase in the outer fuel rod diameter up to 7.2 mm leads to a decrease in the maximal temperatures of fuel and fuel rod cladding. Maximal temperatures of fuel and fuel rod cladding do not exceed melting temperatures for silumine and E635 alloy, respectively. Coolant flow velocity for all fuel rod diameters does not exceed 10 m/s and there is no boiling crisis. [ABSTRACT FROM AUTHOR]

Recently, many studies on nuclear-powered ships as replacements for ships using fossil fuels are attracting worldwide attention. Among fourth-generation reactors, molten salt reactors (MSRs) are a suitable candidate as a power source for nuclear-powered ships because of its high efficiency and safety features. In addition, a small modular reactor (SMR) design using MSR can be developed by adopting the supercritical CO 2 Brayton cycle (SCBC) as a power conversion system. The SCBC is an appropriate system for MSR-powered ships in terms of its operating temperature and compact size. In this study, we compare the design considerations of SCBC with the design requirements of a reference ship and reactor. Sensitivity studies were conducted with optimization variables, i.e., flow split ratio and pressure ratio, to determine the optimal design of the SCBC. An optimal cycle efficiency of 47.78% was obtained when the flow split ratio was 0.7 and the pressure ratio was 2.94. Finally, the SCBC was compared to a reference design using a pressurized water reactor based steam Rankine cycle. The SCBC was evaluated to be a better design for an MSR-based ship, with approximately 12% higher cycle efficiency than the reference design. • Molten salt reactor and supercritical CO 2 Brayton cycle (SCBC) was combined. • The reference ship and reactor were selected for the basic requirements of SCBC. • Recompression cycle layout was adopted for SCBC. • Sensitivity study with the optimization variables was conducted. • The efficiency of optimal SCBC design was higher than a steam turbine system. [ABSTRACT FROM AUTHOR]

The Small Modular Reactors (SMRs) use conventional solid fuels. To increase the safety of SMRs, Dual-Cooled Fuel (DCF) can be used. The DCF with two heat transfer surfaces improves heat removal from the fuel. For the first time, this study investigated the use of DCF in a new fuel assembly configuration for the NuScale reactor. Important neutronic and thermohydraulic (N–Th) parameters are obtained. Finally, the optimal DCF rod in a 12 × 12 configuration for the NuScale reactor has been presented. For this purpose, initially, a DCF assembly in a 12 × 12 configuration was designed for the reactor. The fuel mass, enrichment of 235U, and dimensions of fuel assemblies and the core are kept fixed. Then, 32 types of fuel rods with different internal radii were proposed for this assembly. Each of these fuels was placed separately in the assembly and simulated by neutronic codes under cold-clean and full power conditions. Also, the designed fuels were simulated using Computational fluid dynamics (CFD) software, and important thermohydraulic parameters were obtained. Also, to select the optimal fuel rod, the Artificial Neural Network (ANN) and Genetic Algorithm (GA) coupled method is used. To investigate the optimal DCF effects on N–Th parameters in the NuScale reactor, the reference reactor was also simulated and compared with the results of optimal DCF. As one of the most important results of this study, the fuel rod with an internal radius of 0.413423 cm was introduced as the optimal fuel. Also, optimal fuel reduces the temperature and pressure drop, and increases the Departure from Nucleate Boiling Ratio (DNBR). [ABSTRACT FROM AUTHOR]

Fossil fuel-based economies must undergo a deep transition for complete decarbonization. To this end, it is widely recognized that all economies must move towards the electrification of energy end uses. Even though there is a part that cannot be electrified, at least at affordable costs; clear examples are heavy road transport, maritime and air transport, and some industrial processes. As a result of the limitations of electrification in certain energy end uses, the potential use of hydrogen as an alternative energy carrier has been examined in recent decades. Hydrogen is seen as a viable option for the medium to long term. Specifically, efforts are being made in the case of the Canary Islands to move towards a carbon-free energy mix. In fact, the aim is to advance the economy decarbonization by 10 years over the date foreseen for both Europe and the rest of Spain. To this end, the HOMER program has been used to analyze the possibility of producing the necessary hydrogen in a scenario applied to the Canary Islands in which the uses of this energy vector have been maximized. This research considers two possible scenarios, together with a sensitivity analysis under the different uncertain conditions associated with the used technologies. The results have been the estimation of hydrogen production costs for high demands, about 230,000 tH 2 /year, in 2040 for an isolated archipelago under two totally different generation scenarios. The first scenario is totally renewable, while the other is based on nuclear generation by means of high-temperature SMRs. The results show, that it would be possible to produce at a cost in the range of 1 €/kgH 2 using nuclear technology and around 4 €/kgH 2 using renewables, with uncertainty cost ranges of around 40%, i.e. costs between 0.85-1.48 and 3.29–4.99 € kgH 2 respectively. [ABSTRACT FROM AUTHOR]

Small modular reactors (SMRs) are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR) has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS) can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.