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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.

A small modular reactor (SMR) offers a promising means to produce carbon-free power. Many SMR designs are intended for multiple module operation, whereby several units are operated as a single plant. As such, the output from the plant can be increased or decreased via regulating the individual units. In order to benefit from such operational flexibility, individual units must be properly controlled, as there are direct and indirect couplings among their dynamics. This research examines such a situation and analyzes the dynamic behavior of a group of pressurized water reactor-type SMRs in an islanded power system under the load-frequency control framework. The detailed models of the SMRs and the associated components and subsystems are developed, and effective control systems are designed to investigate the performance of this multimodule topology in supplying uninterrupted power to the loads under different operating scenarios. The performance of the system is evaluated through simulations in PSCAD/EMTDC using a benchmark 9-bus power system.

In this study, lattice physics calculations were carried out to evaluate the reactor physics characteristics of different advanced fuel lattices cooled with 7LiOH/NaOH or FLiBe and moderated extern...

The small modular reactor (SMR) is considered one of the important energy sources for the realization of the de-carbonated society, especially SMR types that have 10 MW or less thermal power, calle...

Small modular reactors (SMRs) are currently framed as a clean energy innovation with a vital role to play in decarbonizing power production. We review two popular sustainability transitions framewo...

In order to ensure safety in a nuclear power plant, operation and protection systems must take into account safety parameters, whether to guide operators or to trip the reactor in emergency cases. ...

Small modular reactors (SMRs) are suitable for deployment in isolated underdeveloped areas to support highly localized microgrids. In order to achieve almost autonomous operation for reducing the cost of operating personnel, an autonomous control system with decision-making capability is needed. In this paper, a decision-making method based on Bayesian optimization algorithm (BOA) is proposed to explore the optimal operation scheme under fault conditions. BOA is used to adjust exploration strategy of operation scheme according to observations (operation schemes previously explored). To measure the feasibility of each operation scheme, an objective function that considers security and economy is established. BOA attempts to obtain the optimal operation scheme with maximum of the objective function in as few iterations as possible. To verify the proposed method, all main pump powered off fault is simulated by RELAP5 code. The optimal operation scheme of the fault is applied, the transient result shows that all key parameters are within safe limits and SMR is maintained at relatively high power, which means that BOA has the decision-making capability to get an optimal operation scheme on fault conditions.

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the Thot and Tcold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactor (SMR) technology. This paper discusses what an SMR is and what potentially makes them different from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges (DIS 16-04) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, the role of the CNSC Vendor Design Review process is described as well as other readiness activities undertaken by the CNSC that are helping to ensure that the CNSC will be ready to accept and evaluate a license application for an SMR.

Thorium fuel is presently a globally known future nuclear fuel alternative, having good neutronic, physical and chemical properties in addition to its spent nuclear fuel characteristic proliferation resistance. This research focused on the neutronic and safety parameters of thorium‐uranium mixed oxide fuel cycle, utilising three fissile enrichment zones, a departure from the conventional single enrichment. The aim was to determine the range of three fissile zones adequate for thorium‐uranium fuel cycle; investigating the performance efficiency of the fuel neutronic and inherent safety parameters in response to temperature differentials, which determines the viability of the fuel and core composition. Use was made of the MCNPX 2.7 code integrated with the CINDER90 fuel depletion code for steady‐state and burn‐up calculations. The keff, moderator temperature coefficient (MTC) and fuel temperature coefficient (FTC) of reactivity are affected by the range of fissile enrichment and fuel temperature which decreased with their respective increases. The MTC for all the moderator temperatures was within 0 to −40 pcm/K design value for UO2 fuel. Similarly, the FTC was within −3.5 to −1 pcm/K design value for all the fuel temperatures except after 2000 days, where a positive reactivity feedback was introduced. At ~86 MWd/kgHM single discharge burn‐up, the result shows that ~90% of the initial fissile load was utilised for energy production at the normal reactor operating temperature (600 K) with a slight reduction at higher fuel temperature. The total fissile inventory ratio (FIR), 233U/kg‐232Th and 239Pu/kg‐238U inventory ratios were significantly large and increased with burn‐up. It is remarkable that the FIR and the 233U/kg‐232Th inventory ratio did not reach conversion equilibrium until exit burn‐up. The large percentage fuel utilisation supports the advantage of fissile enrichment zoning in a thermal nuclear reactor core, making the chosen novel three fissile enrichment zones for thorium‐uranium fuel cycle reliable.

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 SMART (System-integrated Modular Advanced ReacTor) is an integral-type small modular reactor developed by KAERI (Korea Atomic Energy Research Institute). This paper discusses the development of a dynamic equivalent model of the SMART control rod drive mechanism that can be efficiently utilized for complicated analysis during the design of the SMART. A semi-empirical approach is used to develop the equivalent model; that is, the equivalent model is defined analytically and verified empirically. Two types of tests, dynamic characteristics tests and seismic loading tests, are conducted for the development and verification of the dynamic equivalent model, respectively. Acceleration response spectra from the seismic analysis based on the developed equivalent model show good agreement with those from the seismic loading tests.

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.

Small modular reactors (SMRs), an emerging nuclear power plant technology, are suitable for large grids as well as remote load centers and offer load following and frequency response capabilities. While the SMRs have expectedly higher response rates, detailed dynamic models including reactor dynamics are necessary for power system dynamic studies. This paper presents a dynamic model of an integral pressurized water reactor (iPWR)-type SMR, modeled in Siemens PTI PSS/E, to assess the contribution of the reactor to the power system dynamics. The proposed SMR model mimics the heat generation process and subsequent heat transfer process with the inclusion of the reactor core based on point kinetics, primary coolant based on natural circulation, and a simplified three lump representation of the steam generator. Controllers are designed to operate the turbine valve and reactor control rod in closed loops. The SMR model is integrated with the modified turbine-governor system and a power system study is conducted. Results show the power system and internal reactor responses when subjected to electrical demand variations of 20 ${\%}$ rated electrical output (REO) with a valve rate limit of $\pm \text{80}{\%}$ REO/min.

This paper provides information on the issues that impact the adoption of Small Modular Reactors (SMR) in microgrids as additional resources. It summarizes the technical and economic characteristics and financial model cases. It also discusses technical issues related to SMR installations and their operation and control in load following and frequency regulation.

The impacts of various ATF (accident tolerant fuel) claddings on neutronic performances of pressurized water reactor fuel assembly and soluble‐boron‐free small modular reactor autonomous transportable on‐demand reactor module (ATOM) are investigated. There are two ATF cladding concepts which are evaluated here: (a) Coating Zircaloy‐4 cladding with thin layer of Cr or Cr alloys; (b) High‐strength and oxidation‐resistant claddings: SS‐304 and FeCrAl. A minor modification of burnable absorber loading in the ATOM core is proposed to adopt the selected ATF cladding and maintain the core performances.

Some energy consumers require power on an anytime, all-year-round basis with a high level of certainty, including defense installations, isolated communities, and some industrial processes. For the...

To enable the design of a light water small modular reactor, the boiling flow inside a helical coil steam generator has been simulated with the two-fluid model in Nek-2P. Nek-2P is the multiphase b...

Canada has attracted specific interest from developers of nonwater-cooled small modular reactor (SMR) technologies, including concepts based on high-temperature gas-cooled reactors (HTGRs). It is anticipated that some research and development (R&D) will be necessary to support safety analysis and licensing of these reactors in Canada. The Phenomena Identification and Ranking Table (PIRT) process is a formalized method in which a panel of experts identifies which physical phenomena are most relevant to the reactor safety analysis and how well understood these phenomena are. The PIRT process is thus a tool to assess current knowledge levels and (or) predictive capabilities of models, thus providing direction to a focused R&D program. This paper summarizes the results of a PIRT process performed by a panel of experts at Canadian Nuclear Laboratories for a limiting or “worst-case” accident scenario at a generic HTGR-type SMR. Suggestions are given regarding the highest priority R&D items to support severe accidents analysis of these reactors.

Design-basis accidents such as control rod ejection accident (REA) must be evaluated as part of a deterministic safety analysis for any reactor licensing process. In the present work, the behavior of a newly developed boron-free light-water small modular reactor (LW-SMR) core under REA-conditions is studied at the beginning of life using the coupled code PARCS/SUBCHANFLOW. The developed LW-SMR core can produce 330 MW th and contains 57 fuel assemblies. Generally, the REA happens due to a mechanical failure of the control rod drive mechanism housing such that the pressure difference between the primary-system and containment ejects a control rod assembly completely out of the reactor core. In this work, the REA analysis is performed at both critical HZP and HFP conditions. The impact of fuel-cladding gap-heat transfer model during the REA is also analyzed. This investigation showed that there is a high margin against fuel and cladding melting. In addition, the maximum fuel enthalpy is found to be far from fuel-cladding failure threshold. The performed simulations have shown that there is a significant impact of the models used to describe the fuel-cladding gap-heat transfer on the prediction of key safety parameters such as the maximum fuel enthalpy.

Criticality search is one of the important aspects to perform accurate lifecycle analysis of nuclear reactor in Monte Carlo code, in which the geometry criticality search is particularly important for some kinds of reactor systems, such as BWR and so on, to use control rods to adjust the criticality and power in lifecycle. To solve the Monte Carlo geometry criticality search problem correctly and efficiently, the Universe Based Geometry Criticality Search (UBGCS) capability along with the Non-Converged Sub-Iteration (NCSI) acceleration method is newly developed and implemented in reactor physics Monte Carlo code RMC. The UBGCS capability relies on the multi-layer universe geometry description structure in RMC and could perform eigenvalue criticality search not only to single or several surfaces, but also complex cells and universes combination. By using the NCSI method, the RMC’s UBGCS capability could achieve accurate and efficient Monte Carlo criticality search for complex problems. Based on the newly developed UBGCS capability and other advanced techniques in RMC, the lifecycle analysis of a nuclear heating small modular reactor concept NHR200-II is taken out to have a better understanding of the reactor physics characteristics of this reactor. NHR200-II is a new 200 MWt commercial size heating reactor using control rods as power regulating system developed by INET, Tsinghua University according to the new energy and environment requirements in China. In this article, the first detailed Monte Carlo model of NHR200-II has been build including different types of complex fuel assemblies, cruciform control rods, full core structures and so on. The cold zero power and hot zero power state of the NHR200-II reactor are analyzed based on the detailed model. Also using the Monte Carlo reactor physics/thermal hydraulics and UBGCS capabilities in RMC, the full core pin-by-pin burnup calculation of NHR200-II are performed and analyzed.

A call for flexible power generation for a wide range of applications is gradually increasing, making Small Modular Reactor (SMR) a heat topic, especially for marine power. Unlike normal reactors, marine reactors demand a long lifetime up to 15 years without refueling, a high power-density and also a higher reliability and security. The standard PWR cannot achieve this goal. Here, a new 13 × 13 assembly loaded with Accident Tolerant Fuels (ATFs): U3Si2 – FeCrAl system, is proposed and analyzed on an assembly-level. U3Si2 is one kind of ATFs with high heavy metal density and thermal conductivity. It has a lower parasitic absorption. FeCrAl has a great radiation resistance, corrosion resistance and thermodynamic properties although with serious neutron penalties. It is currently the fastest researched accident tolerant cladding material. The fuel enrichment of this assembly is raised to 13%, prolonging the burnup to up to 95,000 MWD/tU, which means that the reactor can survive for more than 15 years at its rated power density. Using 15 B4C control rods can help achieve the beginning-of-life cool shutdown (K

Small modular reactors (SMR) are in great demands in the near future and the lead-based fast reactor is one of the most important directions in the development of SMR. In this paper, core design and analysis of a LEad-bismuth Small MOdular Reactor (LESMOR) was developed and examined to realize the notion of “LESS is MORE”. LESMOR is a 300 MWt pool-type reactor where all facilities are mounted in a reactor vessel. Advanced nitride fuel is adopted, and the spent-plutonium is used as the driven fuel and thorium is used as the fertile fuel. Sub-channel analysis was performed in the assembly design using an in-house sub-channel code SUBAS, and the 11 × 11 scheme with a pitch-to-diameter (P/D) ratio of 1.4 was adopted. The full core neutronic performance was studied and evaluated in this paper. Results show that the reactor can be maintained critical for 20 years and the maximum radial power factor is 1.28. The steady-state thermal-hydraulic performance of the core was analyzed, including the characteristics of the mass flow distribution and the hottest assembly. The peaking fuel cladding temperature is 519.2 °C, and the maximum fuel center temperature is 723.4 °C, both within the temperature limit. This paper provides valuable information in the neutronic and thermal-hydraulic design and analysis of LBE nuclear reactor.

This work presents the safety-related neutronics and thermal-hydraulics features of a reactor core designed to fit into a Small Modular Reactor (SMR) moderated and cooled by light water. The core design process is conducted in an iterative manner to optimize its performance from safety perspective. The optimized core is operated without boron in the moderator and is characterized by enhanced safety features as a result of (1) avoiding concerns related to boric acid induced corrosion of a reactor pressure vessels internals; and (2) eliminating the probability of boron dilution accidents. The optimization studies are performed with an advanced multi-physics simulation tools including both neutronics (deterministic, Monte Carlo) coupled with a subchannel thermal-hydraulics code. With an intensive core design iteration process and through using such multi-physics tools, a reactor core with sufficient thermal margins, a cold shutdown margin, and a reduced power peaking factor is obtained. In this paper, the core design approach, core features, and the verification of the core analysis method are discussed in details. Finally, it has been concluded that the developed and optimized core fulfills all safety requirements and design objectives.

When the small modular reactor (SMR) operates off-grid, the load following demand increases. To solve this problem, a flexible load is used to compensate the electricity load. The flexible load comprises power-consuming systems that can change their load actively, such as district heating, desalination, hydrogen production, and other industrial heat supply systems. Based on the modeling of the nuclear steam supply system, turbine, and flexible load, a simulation platform is established in the MATLAB/Simulink environment. For the load following to be compensated by the flexible load, the key problem is dynamic load allocation, in which the electricity load is divided into a moderate part and fluctuating part. The two parts of the load are undertaken by the SMR and flexible load, respectively. Based on the characteristics of different flexible loads, the changeable flexible load and fluctuating flexible load are presented. Two load following control methods are presented for the two types of flexible loads. The simulation results show that the present method can not only relieve the load following burden of the reactor core but also provide steam with stable quality to the flexible load, thereby improving the flexibility and efficiency of the SMR.

An ex-vessel–type control element drive mechanism (CEDM) is installed on the nozzles of a reactor vessel closure head. However, there has been a demand for locating CEDMs inside the reactor...

At present, no clear guidelines exist for modeling non-water-cooled small modular reactors (SMRs) despite the rising need for high-fidelity simulation tools to support regulators and the industry. ...

Small Modular Reactor (SMR) is a small, compact version of a conventional Nuclear Power Plant (NPP) and holds much promise for the future. Installing SMRs in a region requires a series of carefully planned steps out of which site selection is a critical one. This paper proposes a novel mathematical model for evaluating potential land sites for their suitability of hosting a modular NPP. Most existing decision making tools for NPP site selection rely on qualitative information from the experts. These tools require significant resources and therefore can only be applied to a limited number of selected sites. The proposed model is a Mixed Integer Non-linear Programming (MINLP) formulation which considers a variety of factors like cost, cooling water availability, earthquake risk, etc. to identify best locations for the SMRs in a distributed power system. A case study based on a virtual EIP simulator is conducted to demonstrate the capabilities of the model by finding the optimal locations of modular NPPs. The model offers a preliminary platform for carrying out further extensive studies.

A continuous-energy Monte Carlo neutron transport solver executing on GPUs has been developed within the Shift code. Several algorithmic approaches are considered, including both history-based and event-based implementations. Unlike in previous work involving multigroup Monte Carlo transport, it is demonstrated that event-based algorithms significantly outperform a history-based approach for continuous-energy transport as a result of increased device occupancy and reduced thread divergence. Numerical results are presented for detailed full-core models of a small modular reactor (SMR), including a model containing depleted fuel materials. These results demonstrate the substantial gains in performance that are possible with the latest-generation of GPUs. On the depleted SMR core configuration, an NVIDIA P100 GPU with 56 streaming multiprocessors provides performance equivalent to 90 CPU cores, and the latest V100 GPU with 80 multiprocessors offers the performance of more than 150 CPU cores.

Fully ceramic micro-encapsulated (FCM) fuel has high thermal conductivity and fission product retention capabilities. It can improve the accident tolerance ability of light water reactors (LWRs). Small modular reactors (SMRs) shows a good inherent safety due to their good design and low power density. Combining FCM fuel and SMR may greatly improve the safety of the reactor. In this paper, a SMR model using FCM fuel is built by referring to the Westinghouse Small Modular Reactor (WSMR). Based on this model, we analyzed some accidents associated with SMR using FCM fuel and compare this information with SMR using UO2 fuel. The results indicate that the steady running temperature of FCM fuel is low and that the temperature increases slowly in accidents. The SMR using FCM fuel has good safety features and delay the melting of the core. Under the conditions of station black-out (SBO) combined with failure of core makeup tanks (CMTs) and ADS automatic depressurization system (ADS), the failure time of the core using FCM fuel is 5.49E4 s, which is 7.10E3 s (1.97 h) longer compared with that of UO2 fuel. When considering reinvestment in CMTs, the time limit of FCM fuel is 1.33E4 s (3.70 h) longer compared with that of UO2 fuel, which offers enough time for the restoration of safety system in the accident state. Because of its low temperature, the hydrogen production rate of FCM fuel is reduced compared with that of UO2. However, given the use of zirconium cladding, the final output of hydrogen is the same as or even more than UO2 fuel, and hydrogen elimination devices must be considered.

Civil marine reactors face a unique set of design challenges. These include requirements for a small core size and long core lifetime, a 20% cap on fissile loading, and limitations on using soluble neutron absorbers. In this reactor physics study, we seek to design a core that meets these requirements over a 15 effective full-power-years (EFPY) life at 333 MWth using homogeneously mixed all-UO2 and micro-heterogeneous ThO2-UO2 duplex fuels. In a companion (Part I) paper, we found assembly designs using 15% and 18% 235 U for UO2 and duplex fuels, respectively, loaded into 13 × 13 pin arrays. High thickness (150 μm) ZrB2 integral fuel burnable absorber (IFBA) pins and boron carbide (B4C) control rods are used for reactivity control. Taking advantage of self-shielding effects, these designs maintain low and stable assembly reactivity with little burnup penalty. In this paper (Part II), whole-core design analyses are performed for small modular reactor (SMR) to determine whether the core remains critical for at least 15 EFPY with a reactivity swing of less than 4000 pcm, subject to appropriate constraints. The main challenge is to keep the radial form factor below its limit (1.50). Burnable poison radial-zoning is examined in the quest for a suitable arrangement to control power peaking. Optimized assemblies are loaded into a 3D reactor model in PANTHER. The PANTHER results confirm that the fissile loadings of both fuels are well-designed for the target lifetime: at the end of the ∼ 15-year cycle, the cores are on the border of criticality. The duplex fuel core can achieve ∼ 4% longer core life, has a ∼ 3% lower initial reactivity and ∼ 30% lower reactivity swing over life than the final UO2 core design. The duplex core is therefore the more successful design, giving a core life of ∼ 16 years and a reactivity swing of less than 2500 pcm, while satisfying all the neutronic safety parameters. In particular, one of the major objectives of this study is to offer/explore a thorium-based candidate alternative fuel platform for the proposed marine core. It is proven by literature reviews that the ability of the duplex fuel was never explored in the context of a single-batch, LEU, SBF, long-life SMR core. In this regard, the motivation of this paper is to understand the underlying physics of the duplex fuel and ‘open the option’ of designing the functional cores with both the duplex and UO2 fuel cores.

Small modular reactor (SMR) is a type of advanced reactor depending on its inherent safety, good economic performance, and multipurpose. SMR always adopts the multi-modular layout, which uses several nuclear steam supply systems (NSSSs) to supply steam to a common steam header. A pressurized water reactor (PWR) type multi-modular SMR (MSMR) system is studied in this paper. Due to the operation demands of continuity and flexibility during load following, a proper control strategy and operation scheme of the MSMR must be investigated. A master-slave control strategy is proposed for steam flow sharing control of the MSMR systems. Under the proposed control strategy, one of the NSSS modules is selected as the master control channel, and its feed water flow rate is regulated using the steam header pressure error. The remaining NSSS modules are treated as the slave channels. The feed water flow rate of the master channel is utilized to slave the other channels, which are flow-controlled. Three operation schemes are set according to different operation demands. (1) Equilibrium Scheme (EquS), in which every module undertakes the same load during the load following operation. (2) Non-equilibrium Scheme (NEquS), in which every module undertakes the different loads during the load following operation. This operation scheme manages more modules operate at full power to ensure its economy and operation simplicity. (3) Constant Output Scheme (ConS), which is mainly used during refueling or maintenance process. When one module is disturbed, other modules undertake its load to ensure the total power output constant. Based on these schemes, four typical operation cases are defined and simulated. Through the analysis of the simulation results, the suitable operation conditions for these operation schemes are suggested, and the relevant control strategy is verified.

The analysis of thermal hydraulic phenomenon in tight lattice sub-channels is of importance in pursuing high conversion ratio of Light Water Reactors (LWR). Tight lattice is also a potential choice for reducing the volume of small modular reactor core, which is benefit for improving economy. In this work, an extensive overview of the experimental and numerical analysis for the thermal hydraulic phenomenon in tight lattice sub-channels was conducted, in order to summarize valuable conclusions and results for future work. The critical heat flux (CHF) and turbulent mixing experiments and theoretical models in tight lattice were overviewed systematically. The large scale flow pulsation in sub-channels and simulating methods were analyzed. Finally, the recommendations for future work were discussed.

This article investigates a high temperature prismatic core small modular reactor concept based on the U-Battery. To achieve a long operational time without reloading of the fuel, a high fissile fuel load is required, to provide the required excess reactivity which needs to be compensated to allow safe operational states in all stages of the life of the core. This is achieved through the means of reactivity control methods. In this article we deploy a design methodology for controlling the reactivity in a design based on the U-Battery. The initial excess reactivity within the core was 2000 pcm, implying that the reactivity control systems were required to bring this to unity. To reduce this reactivity, we investigated three main reactivity control methods. Initially the control rod designs which is kept constant. Then a full range of fixed burnable poison designs are evaluated. The final reactivity control method investigated the process of flux spectrum hardening by changing the moderation within the core. This work managed to identify that maximum full power days without any reactivity control as 1364 days. The fixed burnable poisons deployed reduced this lifetime by 62 days due to a reduced neutron economy. However, by removing part of the central reflector the flux spectrum hardens (increases the proportion of fast neutrons) this provides additional bred plutonium provided an additional 62 full power operating days. The meant that the overall design saw no lifetime reduction due to reactivity control devices, without including the use of control rods. Due to the implemented reactivity control systems, the highest point of reactivity has been identified as 1.03, this meant that the control rod position required to reduce this reactivity is limited to the maximum of 50% insertion. Identifying the power profile at this position showed additional loading at the fuel below this point, where additional thermal distribution work will be undertaken.

The small modular reactor market is starting to take shape for the future global energy challenges, with several major players emerging with new technologies. To maximise the potential in this market, the most appealing designs will have to be financially favourable and low risk for investors. Due to its perceived lower risk, this article investigates a high temperature reactor conceptual design proposal, the U-Battery. One of the challenges which high temperature reactors face, is the material selection due to the high temperature in the core and the temperature gradients across the core in addition to the mechanical effects which graphite faces as it ages. The original design is based on using a novel Beryllium oxide reflector. This article investigates the consequences of replacing this novel approach with the more common approach of a graphite reflector. This article aims to provide a comparison between Beryllium oxide and nuclear graphite as a neutron reflector for high temperature reactor. The article emphasises on at what additional costs Beryllium oxide create and if graphite could meet the same performance at lower cost. To deduce a comparison between the materials, a model based around the U-Battery was produced. A neutronic analysis implied that to obtain a similar performance the core would be require 23% larger radial reflectors and 75% larger axial reflectors made if graphite rather than that of Beryllium. However, the cost of the graphite reflected core would be nearly half of that of a Beryllium design. The neutronic analysis demonstrates that the graphite design can be an efficient alternative to the Beryllium reflector creating even slightly better power distributions.

A small modular reactor (SMR) is a promising candidate for future nuclear energy; therefore, many organizations are developing SMRs. Some SMRs have a power output higher than 100 MW(electri...

Nuclear power can solve the energy trilemma of supplying baseload, clean and affordable power. However, a review of nuclear power plant (NPP) builds show mixed results, with delays in Finland and in the US offset by successes in China, South Korea and the UAE. In the West, financing for new builds has been difficult in the face of a deregulated energy market, billion-dollar upfront investments, long build times and in the case of the US historically low gas prices. We explore how the nuclear industry is innovating in facing these challenges through a review of nuclear power developments in the past, present and future. Early developments in nuclear power in the 1950s resulted in a variety of designs, out of which the pressurised water reactor (PWR) became dominant for its compactness and overall economy. Over the next 10 years, several PWR-based small modular reactor (SMR) designs are expected to come online within an eight-year timeframe. Their modular construction and fabrication in a controlled factory setting aims to shorten build times from 8 to 3 years. However, the lack of established regulatory approval pathways may be a time-limiting challenge that needs to be overcome by the first fleet of SMRs. The passive safety and a smaller fuel loading of SMRs will allow them to be deployed at more potential sites, including brownfield replacements of old coal-fired power plants or power unconventional, remote or islanded grids. Some SMRs are also designed to load follow which will allow them to work harmoniously with intermittent renewables sources with the promise of an affordable, truly carbon-neutral grid. In the longer term, advanced nuclear reactors in the form of sodium cooled, molten salt cooled, and high temperature gas cooled reactors hold the promise of providing efficient electricity production, industrial heat for heavy industry as well as the generation of hydrogen for synthetic fuel.

Multi-unit small modular reactor (M-SMR) is a nuclear power plant with multiple nuclear steam supply systems (NSSSs) connected to a single shared secondary loop system. Unlike single SMR, strong coupling characteristics of M-SMR will affect safe, smooth and steady operation under both fault and unbalanced load conditions. In order to improve fault-tolerant operation performance, an effective coordinated control strategy (CCS) is required. In this paper, a novel fuzzy logic based CCS with a M-SMR coordinating controller has been proposed, which utilizes two sets of local controllers for each subsystem. In order to meet the operation goal, the coordinating controller generates coordinated control actions using the fuzzy logic approach to coordinate the local controllers and to perform specific control actions. In this paper, we designed two different local controllers to ensure an excellent performance. The first is a back-propagation neural network (BPNN) based PID controller with capability for floating control and the second is a fuzzy controller with capability for quick regulatory control. The control strategy is designed to ensure operation safety of the reactors and to guarantee consistent and stable steam output of the secondary side under fault and unbalanced load conditions. To evaluate the fuzzy logic based CCS, we simulate faults in the 1# NSSS main coolant pumps (MCPs) which create considerable instability to the coordinated control. Simulation results show smoother and steadier operation performance, compared to the BPNN-based PID control strategy.

The Small Modular Reactor (SMR) concept is designed such that it will solve some of the construction problems of large reactors. SMRs are designed to be “shop fabricated and then transported as modules to the sites for installation” (IAEA, 2018). As a consequence they theoretically have shorter build schedules and require less capital investment (Locatelli et al., 2014). Factory built modules can also increase safety and productivity, due to higher quality tools and inspection available. A literature review has highlighted substantial work has been undertaken in the research, development and construction of different types of reactors and reactor modules but the design of balance of plant modules has not been extensively researched (Wrigley et al., 2018). The focus of this paper is a case study for balance of plant modules in a light water reactor which also could have applications to other reactor types. Modules that are designed for factory build and transport may be built in a standardized module approach. By maximizing module size for transport, this maximizes work offsite, to achieve the cost and schedule savings associated. A design method needs to be developed to help support this approach. To enable this, a three step method is proposed: group components into modules, layout the modules and arrange components inside the modules. The Shearon Harris nuclear power plant was chosen for its publically available data. A previous study on this plant used matrix reordering techniques to group components and heuristically assign them to large modules, built for construction in an assembly area on site, highlighting a potential capital cost savings of 15%. This paper utilizes the same allocation of components to modules as the previous study but aims to undertake the challenge of how balance of plant modules should be arranged. The literature review highlighted that although the facility and plant layout problem has been extensively researched, mathematical layout optimization has not been applied to nuclear power plants. Many techniques for layout optimization have been developed for facilities and process plants however. The work in this paper develops an optimization model using a genetic algorithm for module layout and allocation within a nuclear power plant. This paper analysed two configurations of modules, where balance of plant modules are located on either one or two sides of the nuclear island. The objective function was to minimise pipe length. In the original research, where the plant was configured for assembly on site, the balance of plant modules are located around three sides of the nuclear island. The objective function was calculated at 14,914. As the distances are calculated rectilinearly, this number would be higher in reality as pipework has to be routed around containment. The optimization reduced the objective function by 33.9% and 37.8% for the three and four floor layouts respectively when balance of plant modules are located on two sides of the nuclear island. Furthermore, when modules are located on one side of the nuclear island, the objective function was reduced by 45.4% and 46.1% for three and four floor layouts respectively. This will reduce materials used, reduce build time and hence reduce the cost of a nuclear power plant. This method will also save design time when developing the layout of modules around the plant.

It has been argued that risk and performance-based approaches to licensing would be appropriate for Small Modular Reactors (SMRs) because their risk profiles differ from large-scale reactors. This is based on several factors including their limited electrical capacity of 300 MW, the below grade reactor vessel, and passive safety features. One design feature that can significantly reduce accident severity is the larger lateral surface area-to-volume (A/V) ratio of SMRs. Following a nuclear accident, this larger A/V ratio can increase the removal of radioactive particles due to natural phenomena compared to large light water reactors (LWRs). To quantify the improvements in safety, this work estimates the airborne radioactivity within containment and environmental dose exposure in a post-accident scenario for an advanced Generation III+ LWR (AP1000), a representative Generation II LWR (Surry), and an SMR. On average, the AP1000, Surry, and SMR produces 139, 153, and 104 curies/ft3 (182, 200, and 136 terabecquerels/m3) 75 min after a Loss-of-coolant-accident (LOCA). Using Monte Carlo simulations, the SMR produces less radioactivity per volume in containment than the AP1000 and Surry 84% and 96% of the time, respectively. On average, the AP1000, Surry, and SMR produces 84, 106, and 7 thousand curies/ MW th (3.1, 3.9, and 2.5 petabecquerels/ MW th ) 75 min after a LOCA. The larger A/V ratio of the SMR plays a substantial role in reducing the radioactivity. While it is expected that the SMR would have a lower levels of radioactivity compared to the AP1000 and Surry, the SMR produces less radioactivity after normalizing by thermal reactor power and containment volume. With respect to environmental dose exposure, the US Environmental Protection Agency 1–5 rem (0.01–0.05 sieverts) Protective Action Guide (PAG) limits for whole body exposure is not exceeded at the 10-mile (16.1-km) EPZ using the mean estimates for the AP1000 and Surry. The iPWR does not exceed the 1 rem (0.01 sieverts) lower PAG limit for whole body exposure at the 5-mile (8-km) EPZ using the mean value. These findings can be used in conjunction with the improved analytical methods, found in the SOARCA study, to provide accurate and realistic estimates for exposure. This will help create a pathway to develop a regulatory basis for technology-neutral, risk-based approach to EPZs for iPWRs.

Several small modular reactor (SMR) designs are emerging, but only the CANDU® SMR and a couple of Indian designs incorporate the familiar features of the larger CANDU reactors. This paper shows that while the CANDU-reactor concept did not seem to receive wider attention among SMR designers, it has influenced a few. The paper discusses how the CANDU-reactor operating experience can aid in the construction and operation of some SMRs. For example, the concept of passive reactor shutdown by draining the moderator, which was utilized in the early Pickering A units, is adopted in the Copenhagen Atomics Waste Burner; a molten slat (LiF-ThF 4) heavy-water moderated reactor. The heavy-water and lithium in this salt produce tritium and can benefit from the CANDU-reactor experience in handling tritium. The online refueling of CANDU reactors, their large heat sinks, and seamless configuration are also reflected in SMR designs.

Designing a novel scaled modular test facility as a part of an experiment for condensation heat transfer (CHT) in small modular reactors (SMRs) is the main focus of this study. This facility will provide data to evaluate models' scalability for predicting heat transfer in the passive containment cooling system (PCCS) of SMR. The nuclear industry recognizes SMRs as future candidates for clean, economic, and safe energy generation. However, licensing requires proper evaluation of the safety systems such as PCCS. The knowledge gap from the literature review showed a lack of high-resolution experimental data for scaling of PCCS and validation of computational fluid dynamics tools. In addition, the presently available test data are inconsistent due to unscaled geometric and varying physics conditions. These inconsistencies lead to inadequate test data benchmarking. To fill this research gap, this study developed three scaled (different diameters) condensing test sections with annular cooling for scale testing and analysis. This facility considered saturated steam as the working fluid with noncondensable gases like nitrogen and helium in different mass fractions. This facility also used a precooler unit for inlet steam conditioning and a postcooler unit for condensate cooling. The high fidelity sensors, instruments, and data acquisition systems are installed and calibrated. Finally, facility safety analysis and shakedown tests are performed.

Use of flexible and dispatchable generation from the small modular reactors (SMRs) in combination with the nondispatchable generation from renewable energy systems (RES) can be an effective alternative to pursue the mandate of replacing the fossil-fuel based electricity with the carbon-neutral energy systems in the remote microcommunities. This paper evaluates the feasibility of SMRs' flexible operations in microcommunities with the photovoltaic (PV) generation as a case study. Considering the design limits of SMRs for (a) the range of net change in electrical power output and (b) the ramp rates of net change in turbine power, a power system study is conducted to cover the three aspects of flexible operations, namely: (1) Planned load-following, (2) Unplanned load-following, and (3) Frequency regulation. A generic governor model in power system simulator for engineering (PSS/E), a power system transmission and planning software, is adapted to incorporate the operating limits of the reactor for the dynamic simulation. The multitimescale approach, combining (a) steady-state time-series power flow analysis and (b) dynamic simulations with high-resolution solar irradiation datasets, is proposed to assess the implications of SMR's design limits. The results obtained on an existing remote feeder with three sets of operating limits—namely, the conventional, advanced and extreme limits of ramp rates juxtapose the SMRs' performance, given the challenging operating conditions with PV generation in remote locations. The results indicate that the SMR under study can accommodate the highest permissible PV penetration obtained by the hosting capacity analysis of the feeder under the clear sky conditions. However, dynamic simulations with the extreme PV variabilities show that the PV penetration level should be further limited so that the maximum deviations in SMR power levels stay within 40% of its rated capacity. SMR provides adequate frequency support for the PV penetration of up to 50% of the feeder maximum demand in this study.

A novel re-optimization of fuel assembly and new innovative burnable absorber (BA) concepts are investigated in this paper to pursue a high-performance soluble-boron-free (SBF) small modular reactor (SMR), named autonomous transportable on-demand reactor module (ATOM). A truly optimized PWR (TOP) lattice concept has been introduced to maximize the neutron economy while enhancing the inherent safety of an SBF pressurized water reactor. For an SBF SMR design, the 3-D centrally-shielded BA (CSBA) design is utilized and another innovative 3-D BA called disk-type BA (DiBA) is proposed in this study. Both CSBA and DiBA designs are investigated in terms of material, spatial self-shielding effects, and thermo-mechanical properties. A low-leakage two-batch fuel management is optimized for both conventional and TOP-based SBF ATOM cores. A combination of CSBA and DiBA is introduced to achieve a very small reactivity swing (

The understanding of the Natural Circulation (NC) mechanism and its flow regimes can be considered as a help to achieve a higher safety level and lower costs by reliance on this mechanism. NC Flow Map (NCFM) envelope diagram as an engineering tool for evaluating the NC Performance (NCP) has been attained from different test facilities database. Some Small Modular Reactors (SMRs) with their interesting safety and economic features have used the NC mechanism as their main primary flow drive mechanism. Uncertainty analysis of NCP of a SMR with NC flow drive mechanism, according to the NCFM envelope diagram, is the main objective of the current study. For this purpose, the NuScale reactor with helical coil steam generator at the normal operation NC mode has been modelled using RELAP5/Mod3.3 and the differences of the results in comparison to the designer’s data are in the acceptable range. Then an uncertainty quantification for several sensitive parameters of NuScale has been performed according to the Code Scaling, Applicability and Uncertainty (CSAU) method and NCFM uncertainties for this reactor have been derived. The NuScale NC uncertainties during the normal operation have been mapped to the NCFM envelope to evaluate the NCP of this reactor in high powers. The mass flow rate to power as an index parameter has been used in NCFM and the value of this parameter is evaluated in the range of 3–4 ( Kg MW . s ). This range shows the better performance and good design of NuScale reactor for normal operation at the NC condition in comparison to the typical PWRs with high power NC.

A supercritical water-cooled reactor (SCWR) adopts a once-through direct cycle, which is compatible with a small modular reactor class (SMR) plant system. The core is cooled by supercritical light water, which does not exhibit phase change, but undergoes large temperature and density changes. A super fast reactor (Super FR) is a fast reactor type concept of SCWR. Unlike other SCWR core concepts, it adopts the single coolant pass flow scheme, in which the coolant passes the core only once from the bottom to the top without any reverse flows or preheating stages. In the meantime, reducing the core size tends to increase the core power peaking and reduce criticality. Therefore, the key issues with the small Super FR core design is reducing the core power peaking and achieving high average core outlet temperature with the single coolant pass scheme. This study aims to highlight the design issues through conceptual core designs of SMR class Super FR. To evaluate the core characteristics, three-dimensional coupled core calculations are carried out. The proposed design with small fuel assemblies, which are equivalent to those of boiling water reactors, attains a high core average outlet temperature of about 500 &deg, C, which is compatible to that of typical large SCWR core design.

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&rsquo, s performance. It was found that nonuniform diameters promote velocity changes that affect the natural circulation flow rate. Moreover, the reactor&rsquo, s temperature distribution depends on the steam generator tube pitch. Therefore, the hydraulic diameters of the reactor&rsquo, 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.

Investment appraisal methods based on real options are gaining popularity in academia, while the adoption by practitioners is still infrequent. Having in mind practitioners, the method is designed to be: conceptually easy to understand, based on realistic hypotheses and data available, able to provide quantitative indications and strategic guidelines. The method is based on a systematic simulation of several scenarios generated according to exercise thresholds of relevant investment parameters. An exercise threshold gives to the investors the exercise right of taking some decision, for instance building a power plant. An exercise threshold is therefore a rule to decide whether to exercise or not a certain option on the basis of the values of one or more state variables. Consequently, the probability distribution of the Net Present Value (or analogous indicator) of the investment is a function of the state variables and the exercise threshold. Systematically changing the exercise thresholds allow to (A) establishing the “real option value” and (B) calculate relevant indications about when and on which type of plant to invest. The method is presented in detail and applied to a case study assessing the investment appraisal of: coal plant, gas plant, large nuclear reactor and small modular reactor.

This article presents the comparison of two reduced moderation small modular reactor concepts with heavy water coolant. Two reduced moderation small modular reactors, RMSMR-Th and RMSMR-MOX, are proposed for the sustainable utilization of nuclear resources. The design concepts are established on modifications of the well-experienced pressurized water reactor technology. Tightly packed lattice and heavy water coolant are employed to yield a hard neutron spectrum, which proved advantageous for increasing the conversion ratio, as well as lowering the burn-up reactivity swing between beginning of cycle and end of cycle. Thorium–uranium dioxide fuel and MOX fuel are compared using the same core arrangement, and the small modular reactor concept is adopted to reduce void coefficients. Radial blanket region and axial blanket region are adopted to enhance the fissile breeding, and a three-zone fuel configuration is adopted to flatten the power distribution. Core burn-up calculations were carried out to investigate the available cycle length, the conversion ratio, the power peaking factors, reactivity coefficients, etc. Light-water–based thermal–hydraulic models were employed to examine the safety features of the concepts. Numerical simulations indicate that both RMSMR-Th and RMSMR-MOX can sustain the power generation of 100 MWe by 7 years without refueling. Compared with thorium–uranium dioxide fuel, MOX fuel is helpful in reducing the burn-up reactivity swing, increasing the conversion ratio, and increasing the minimum departure from nuclear boiling ratio value. However, the positive void coefficient becomes a problem making RMSMR-MOX less attractive than the RMSMR-Th concept.

With a steady rise in power demand in the remote communities in Canada, utilities are looking for new options to provide a reliable supply of electricity. While distributed generation is a promising option, scaling and firming up the capacity of distributed generators is essential. Alternatively, small modular reactors (SMRs) can be used as a prime local source of electricity for remote feeders provided they are flexible enough to respond to the fluctuations in demand. Electrical energy storage (EES) can be used as a buffer to absorb fluctuations in demand and generation, and as a critical back-up for the SMR on-site power supply system by replacing the diesel-generator sets. The synergy of SMR-EES-distributed generation can be an all-inclusive alternative with win-win situation for both the utility and remote communities. This paper discusses the technical feasibility of the proposed synergy using an example of an existing remote feeder in Saskatchewan, Canada. The integral pressurized water reactor is considered along with the photovoltaic (PV) generation in an existing remote feeder in Northwest Saskatchewan to estimate the plant load factor (LF) of the SMR with and without the PV generation and EES. The results quantify the benefit of having EES to support the SMR in hosting more PV generation in remote communities. EES when used in support of the SMR to host 60% PV penetration, the plant load factor improves by as much as 5%.