Search

Currently, little is known about the characteristics of oxide films on Ti alloys exposed to the secondary circuit water environment of small modular reactors (SMRs). In this study, multiple microstructural characterization techniques were employed to meticulously characterize the oxide film characteristics of the TA16 alloy after 6000 h and 8000 h of exposure to a simulated SMR secondary circuit water environment. The results indicate that the oxide films on the 6000 h specimen are composed of a microcrystalline anatase TiO2 interior layer and a coarse-grained anatase TiO2 exterior layer, while the oxide film on the 8000 h specimen typically consists of a microcrystalline anatase TiO2 inner layer, an NiFe2O4 middle layer, and a coarse-grained anatase TiO2 outer layer. As the corrosion time is extended to 8000 h, the presence of NiFe2O4 oxide particles can be attributed to oxide deposition by the 316L stainless steel autoclave used in the corrosion experiments, which results in the migration of Ni and Fe ions onto the surface of the Ti alloy.

The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48–97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56–70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.

The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive and Discrete Gadolinia/Alumina Burnable Absorber (HIGA) rods for controlling excess reactivity sustainably. Through comprehensive analysis and simulations, the reactivity behavior with varying quantities of HIGA rods is examined, leading to the development of optimized fuel assembly designs. Furthermore, the integration of HIGA rods with integral gadolinia BA rods is discussed to enhance reactivity control and operational flexibility further. This approach utilizes the spatial self-shielding effect of gadolinia for extended reactivity management, crucial for stable and efficient reactor performance. The paper thoroughly addresses core design considerations, including fuel assembly configurations and control rod patterns, to ensure safety and performance in initial and reload cycles. This research advances the development of SBF operation in i-SMR by offering practical reactivity management solutions.

This paper proposes a flooding safety system (FSS) and its operation strategy that can provide long-term safety and effective maintenance for modules of small modular reactor (SMR) and metal containment maintained at dried environment during normal operation. During hypothesized accidents, the FSS re-collects the evaporated steam into the common pool by the condenser installed above the common water pool and provides an emergency coolant for the cavities and auxiliary pools. This study suggested that the condensate re-collection strategy using the FSS can effectively delay the depletion of available water in response to the accidents. Without re-collection, the achievable grace periods ranged from 44 to 1507 days for six-module and one-module accidents, respectively. However, with a full re-collection (ratio = 1.0), the time to total depletion of emergency coolant was estimated indefinite. Even with a partial re-collection ratio of 0.3, a grace period of 83.5 days could be ensured for a six-module transient. This study reported the effectiveness of condensate re-collection and the FSS as an innovative safety management strategy and system. Employing a condensate re-collection strategy with a high re-collection ratio can enhance the long-term safety and effective convenience of SMR operations and maintenance.

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

The Small Modular Reactor emerges as a recommended technology for nuclear power plant (NPP) development in West Kalimantan, particularly at the Pantai Gosong site. NPP demand highly purified water for cooling system, while seawater contains high TDS, so desalination by utilizing heat and power from NPP, offers a viable solution. Additionally, desalination products can supply freshwater to communities within a 10 km radius, including Sungai Raya and Sungai Raya Island Districts. This study aims to determine the techno-economic aspect of desalination cogeneration with SMR-type NPP for West Kalimantan. The methodologies in this research encompass a comprehensive literature review, a qualitative assessment of desalination technology, and a quantitative economic analysis utilizing the DEEP version 5.1. Techno-economic analysis shows that the simple payback period for MED is 12.3 years, while MSF is 13.9 years. Power lost at MED is 5.3 MWe, while MSF is 9.1 MWe. At a temperature of 30.7 °C, the total water cost of the MED is $1.277 $/m3, while MSF is $1.595 $/m3. MED desalination with temperature 70 °C can be effective using residual heat as an energy source. Cogeneration of MED technology can be option for utilization of heat and power in the context of NPP development in West Kalimantan.

The corrosion behavior of a TA16 titanium alloy during exposure in simulated water environment of small modular reactor (SMR) at 320°C for up to 5000 h was investigated by employing dedicated microstructure characterization. The corrosion weight gain curve of the alloy is found to follow a parabolic law with an exponential parameter of n = 0.738. Oxide scales formed on the alloy surface generally show a two-layer structure composed of a relatively compact inner layer (dense oxide layer) and a discontinuous outer layer (microcrystallites). The dense oxide in the inner layer corresponds to anatase TiO2 while the surface microcrystallite layer is mainly comprised of ilmenite FeTiO3. The number and size of the microcrystallites in the outer layer gradually increase with the exposure time. Fe ions from the corroded loop pipe materials should have participated in forming the outer layer. The compact inner layer is thickened gradually with the test time with its thickness growth curve featured by l = 4.245t0.588, which could better reflect the corrosion behavior of the TA16 alloy.

The use of nuclear energy can contribute to achieving positive socio-economic and environmental benefits, but nuclear power plants are one of the most water-intensive industries in the world. The use of Small Modular Reactor (SMR) technologies is increasing due to their interesting advantages such as reduction of construction costs and use in remote areas, which favors distributed generation. Hygroscopic Cycle Technology (HCT) can be of great interest for power generation in nuclear power plants, due to the potential improvement in terms of energy efficiency and water savings. This study presents the benefits of implementing HCT in an existing SMR, the HTR-10, based on the classical Regenerative Rankine Cycle (RRC). The HTR-10 is used to produce electricity and thermal energy for District Heating (DH). Analytical models of both cycles have been developed to compare them in terms of energy production and water consumption. Sensitivity analyses of the influence of the main variables have been performed. The results show that by varying the condensing pressures, the thermal power for DH and the net mechanical power production of the HCT increase up to 2.5 % and 1 %, respectively, with respect to the RRC. The maximum tolerable ambient temperature for the plant with the HCT is 43.12 °C, increasing the availability of the plant and avoiding water consumption between 70000 and 88000 m3/year, depending on the operating conditions. Extrapolation of the results suggests that HCT can improve the energy production of nuclear power plants in a more sustainable way, contributing significantly to the energy transition.

Hydraulic performance and flow resistance tests were performed to confirm the main parameters of the hydraulic instrumentation that can affect the pump performance of the reactor coolant pump. The flow resistance test offers important experimental data, which are necessary to predict the behavior of the primary coolant when the circulation of the reactor coolant pump is stopped. Moreover, the shape of the hydraulic section of the pump, which was considered in the test, was prepared to compare the mixed-flow- and axial-flow-type models, the difference in the number of blades of the impeller and diffuser, the difference in the shape of the impeller blade and its thickness, and the effect of coating at the suction bell. Additionally, five models of the hydraulic part were manufactured for the experiments. In this study, the differences in performance owing to the design factors were confirmed through the experimental results.

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

Nuclear energy sources, in particular the small modular reactor (SMR), are being recognized as low-emission clean energy sources. The development of SMRs has focused on minimizing their carbon footprint while ensuring that energy is generated in a safe and cost-competitive manner, which leads to a worldwide growing interest in SMR deployments. In addition, there is a strong demand for renewable energy sources; however, their output can be influenced significantly by weather and location. Therefore, integration of renewable energy sources with SMRs and an energy storage system, which results in a nuclear renewable hybrid energy system (NR-HES), is required to improve power generation reliability. NR-HES represents a promising energy system solution for off-grid communities and industries, especially those involving remote mining sites. In the present study, an off-grid NR-HES to meet electrical and thermal demands is considered for the Victor mine site in northern Ontario, Canada. Off-grid remote mines in Canada primarily rely on diesel generators. The proposed NR-HES is composed of SMR units with a thermal capacity of 15 MWth each, two wind turbines with a capacity of 2.3 MWe each, a two-tank molten salt thermal storage system, a hydrogen storage facility, and peaking diesel generators. To achieve a near-zero-emission mining site, electric vehicles are proposed in the mining site energy demand structure. A transient thermodynamics model is developed to simulate the NR-HES and generate hourly variation of sustainability performance parameters that include power cycle efficiency, SMR capacity factor, round-trip efficiency of the storage systems, thermal storage state of charging, and greenhouse gas emissions. In addition, an economic analysis is performed to determine the additional annualized cost of the system relative to a baseline site. The simulation has a test matrix of 16 runs with various combinations of design input parameters, namely SMR thermal power, storage duration (e.g., daily, seasonally, or monthly), and thermal storage system volume. Grey relational analysis (GRA) is used as a level-based multi-objective optimization technique to determine the nearest optimal design parameter (e.g., decision variable) combination within the test matrix and thus the optimal configuration. In addition, the GRA ranks the design input parameters based on how significantly each parameter can affect the overall performance of the NR-HES. Results revealed that optimal sustainable perspectives have a significant influence on the overall system parameters and performance. For instance, the results showed that there is an opportunity to achieve a reduction in greenhouse gas emissions by 94.8% with a minimal impact to system annualized cost if the cost and environmental effect objectives were prioritized over other sustainability performance parameters.

The once-through steam generator used in the small modular reactor needs to consider the stability of the outlet steam pressure and steam superheat of the secondary circuit to achieve better operating efficiency. For this reason, this paper designs a controllable operation scheme for the steam pressure and superheat of the small reactor once-through steam generator. On this basis, designs a variable universe fuzzy controller, first, design the fuzzy control rules to make the controller adjust the PI controller parameters according to the change of the error; secondly, use the domain adjustment factor to further subdivide the input and output domain of the fuzzy controller according to the change of the error, to improve the system control performance. The simulation results show that the operation scheme proposed in this paper have better system performance than the original scheme of the small reactor system, and controller proposed in this paper have better control performance than traditional PI controller and fuzzy PI controller, what's more, the designed control system also showed better anti-disturbance performance in lifting experiment between 100% and 80% working conditions. Finally, the experimental platform formed by connecting the digital small reactor with Matlab/Simulink through OPC(OLE for Process Control) communication technology also verified the feasibility of the proposed scheme.

Small modular nuclear reactors (SMRs) are an emerging technology with many potential benefits to cost, manufacturing, and safety over more traditional reactor designs. The NuScale Power Module (NPM) is a pressurized water SMR that uses natural convection to circulate its coolant. Variations of steam Rankine cycles have been the predominant design for power cycles in the nuclear energy space for decades, but ongoing research has suggested promising alternatives to be used in future designs. Previous works have suggested that a transcritical Rankine cycle with an organic working fluid like ethanol or methanol could have superior thermal efficiency compared to comparable steam Rankine cycles for mid-temperature applications. This study presents a design optimization and sensitivity analysis of an organic transcritical Rankine cycle (ORTC) with methanol as a working fluid, with a model of the NPM used as the primary cycle. Multi-objective optimization is used to analyze the trade-off between thermal efficiency and levelized cost of energy (LCOE) for this cycle design. All of the optimal designs identified in the multi-objective optimization show improvements in LCOE compared to the benchmark regenerative steam Rankine cycle at comparable thermal efficiencies, with LCOE being reduced by as much as 19.1% by using the ORTC. In the Pareto front, higher efficiency solutions use regenerative cycle designs, while lower LCOE solutions do not utilize regenerators. A sensitivity analysis of the optimal design points reveals that regenerative design points have greater sensitivity in both LCOE and thermal efficiency to the variation in the design parameters.

This paper presents a high performance burnable absorber named as CIMBA (Cylindrically Inserted and Mechanically Separated Burnable Absorber) for the soluble-boron-free SMR. The CIMBA is the cylindrical gadolinia inserted into the annular fuel pellets. Although the CIMBA utilizes the spatial self-shielding effect of the fuel material, a large reactivity upswing occurs when the gadolinia is depleted. To minimize the reactivity swing of the CIMBA-loaded FA, two approaches were investigated. One is controlling the spatial self-shielding effect of the CIMBA as burnup proceeds by a multi-layered structure of the CIMBA (ML-CIMBA) and the other is the mixed-loading of two different types of CIMBA (MIX-CIMBA). Both approaches show promising performances to minimize the reactivity swing, where the MIX-CIMBA is more preferable due to its simpler fabrication process. In conclusion, the MIX-CIMBA is expected to accelerate the commercialization of the CIMBA and can be used to achieve an optimal soluble-boron-free SMR core design.

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

To meet the energy demand in remote areas of western China, a small modular fluoride-salt-cooled high-temperature reactor will be deployed there. The design of this reactor needs an application target image, including the power, lifetime, size, weight, and environmental restrictions. This paper analyzes the energy demand in northwest China to find the possible application scenarios and regions of this reactor. Then according to typical application scenarios, the power requirements of the reactor and the environment and transportation (size and weight) restrictions that need to be adapted can be obtained. The application target image of this reactor would be: a) the single unit capacity of 50MWe; b) the lifetime of at least 60 years; c) the length of less than 15.5 m, the diameter of less than 3.88 m, and weight of less than 60 tons; d) the adaptability to the severe climate and environment in the west, such as cold winter, deep-frozen soil layer, strong wind, arid environment, and complex terrain; e) the site selection to avoid geological disasters, such as landslides, collapses, torrents, mudslides, and ground fissures.

This paper newly proposes an efficient analytic springback prediction method to predict the final dimensions of bent cylindrical tubes for a helical tube steam generator in a small modular reactor. Three-dimensional bending procedure is treated as a two-dimensional in-plane bending procedure by integrating the Euler beam theory. To enhance the accuracy of the springback prediction, mathematical representations of flow stress and elastic modulus for unloading are systematically integrated into the analytic prediction model. This technique not only precisely predicts the final dimensions of the bent helical tube after a springback, but also effectively predicts the various target radii. Numerical validations were performed for five different radii of helical tube bending by comparing the final radius after a springback.

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO2 and (U, Th)O2 fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO2 fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O2 fuel. The excess reactivity in all FAs with (U, Th)O2 fuel was higher than UO2 fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th)O2 fuel remained positive (16,000 pcm) while UO2 fuel shows a negative value (−6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O2 fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO2 fuel than in (U, Th)O2 fuel except for 236Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for 135Xe were respectively smaller in (U, Th)O2 fuel than in UO2 fuel but may be two times higher in (U, Th)O2 fuel due to its potential longer discharge burn-up.

Reduced-pressure electron beam (EB) plate butt welds were manufactured in two low-alloy pressure-vessel steels, SA508 Gr 3 Cl 1 and SA508 Gr 2, at two thicknesses in both steels, 30 mm and 130 mm. Transient temperatures during welding were recorded using thermocouple arrays. Residual stresses in the as-welded condition and after post-weld heat treatment were measured using diverse methods: neutron diffraction and the contour method at 30 mm thickness; and deep hole drilling and the contour method at 130 mm. Incremental centre hole drilling measurements were performed at 130 mm thickness to better understand near-surface stresses. Weld and heat-affected zone microstructures and microconstituents were evaluated using a combination of hardness mapping, optical microscopy and electron microscopy. The as-welded residual stresses exhibit the characteristic M-shaped distribution for hardenable steels, reaching 500–600 MPa in tension in both steels at both thicknesses. However, the modest changes to the chemical composition and the change in plate thickness both significantly influenced microstructures, mechanical properties and residual stress distributions. These sensitivities underline the need for physically faithful models. This extensive characterisation study enables the development and validation of models that predict the development of microstructures and residual stresses in EB welds in low alloy pressure vessel steels.

Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

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.

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.