Large Reactor Research Articles

The state of progress and associated challenges in the global deployment of small modular reactors

Small Modular Reactors (SMRs) represent an innovative class of reactors with the advantages demonstrated through the reliable history of the current fleet coupled with a unique operational paradigm that may facilitate their deployment. This paradigm refers to new markets: nuclear hybrid energy systems; baseload generation to back intermittent renewable sources or process heat applications; decentralised electricity grid generation, where electricity demand may not support large reactors or the water supply is limited; and construction of single modules at a time, allowing new investors to enter the market with lower investments risks. With the increase in wind and solar generation worldwide, the inherent electricity grid and market fluctuations must be addressed. SMRs offer one potential solution; however, the social, political, and economic challenges must be overcome in regions where nuclear power is already prevalent. Other obstacles exist in new nuclear markets, including the development of a regulatory body and supply chain.

Implication of LOCA characteristics of large PWR and SMR for future development of intelligent nuclear power plant control system

Operation with reduced number of operators is considered as an important operational mode to improve economics of a nuclear power plant (NPP) in the future and it will be more important for the small modular reactors (SMRs). To control NPPs efficiently and safely with a reduced number of operators, an intelligent operation system can be very useful. However, reduced number of operators will result in larger consequence in human error due to simultaneous multi-unit failures are now more probable than the conventional operational mode. Thus, to decrease the probability of human error occurrence for the reduced number of operators, the development of an autonomous operation system with certain level of intelligence is inevitable. The prediction capability of system’s behavior as well as understanding the root cause for the current situation should be an essential function of the designed autonomous operation system. The authors propose to use the current system thermal–hydraulic (STH) analysis code platform to develop such intelligence. As a demonstration of the suggested method, the authors used the STH code to track the accident initiator of both large pressurized water reactor (PWR) and SMR. The discussions on how this method can be useful for future NPP operation and how generic differences in system design actually influence the system behavior are presented.

Use of Dynamic Event Trees and Deep Learning for Real-Time Emergency Planning in Power Plant Operation

An initiating event that disrupts regular nuclear power plant (NPP) operation can result in a variety of different scenarios as time progresses depending on the response of standby safety systems and operator actions to bring the plant to a safe, stable state, or the uncertainties in accident phenomenology. Depending on the severity of the accident and potential magnitude of release of radioactive material into the environment, off-site emergency response such as evacuation may be warranted. An approach that could be used for real-time emergency guidance to support the declaration of a site emergency and to guide off-site response is presented using observable plant data in the early stages of a severe accident. The approach is based on the simulation of the possible NPP behavior following an initiating event and projects the likelihood of different levels of off-site release of radionuclides from the plant using deep learning (DL) techniques. Training of the DL process is accomplished using results of a large number of scenarios generated with the Analysis of Dynamic Accident Progression Trees/MELCOR/Radiological Assessment System for Consequence Analysis (RASCAL) computer codes to simulate the variety of possible consequences following a station blackout event (similar to the Fukushima accident) for a large pressurized water reactor. The ability of the model to predict the likelihood of different levels of consequences is assessed using a separate test set of MELCOR/RASCAL calculations.

Mathematical modeling and simulation of an industrial adiabatic trickle-bed reactor for upgrading heavy crude oil by hydrotreatment process

The mathematical modeling of a trickle-bed reactor at particle and catalytic bed levels for the hydrodesulfurization and hydrodemetallization of heavy crude oil was carried out. The kinetic model parameters were estimated from experimental data. Thermodynamic and transport properties were calculated based on existing correlations and equations of state. Simulations of isothermal small scale and large scale adiabatic reactor were carried out and concentration and temperature profiles were obtained. It was demonstrated that temperature affects the catalyst utilization as the effectiveness factor is diminished at higher temperatures. Effectiveness factors undergo changes as reaction mixtures passes through catalytic bed describing different profiles for HDM and HDS.

The Industrial Production of Dimethyl Carbonate from Methanol and Carbon Dioxide

This work discusses the design of a dimethyl carbonate (DMC) production plant based on methanol and CO2 as feed materials, which are a cheap and environment-friendly feedstock. DMC is a good alternative for methyl-tert-butyl ether (MTBE) as a fuel oxygenating agent, due to its low toxicity and fast biodegradability. Based on the MTBE demand of a general gasoline plant, the annual production capacity of the process design is stipulated to be 86 kt DMC, with a purity of 99 wt%. Three routes are proposed to form DMC: 1) direct synthesis from methanol and CO2, 2) reaction of CO2 and ammonia to urea, which can be converted to DMC with methanol, 3) reaction of ethylene oxide with CO2 to a cyclic carbonate, which can be converted to DMC by transesterification with methanol. From a black box cost analysis based on raw material prices, it is concluded that the ethylene oxide route is the least profitable. Because of higher single-pass conversions found in literature, smaller recycles and easier separations, it is concluded that the urea route would be the most feasible. The required process functions for the urea route have been determined in the conceptual design phase. A detailed design of the most important process operations is made and an overall technical and economic evaluation of the process has been carried out. In the first step of this DMC synthesis, urea is produced from carbon dioxide and ammonia with the ACES21 process. After separation and purification steps, urea is fed to a reactor with methanol (150 °C, 20 bar), where methyl carbamate (MC), an intermediate of DMC production, and ammonia are formed in the absence of a catalyst. Subsequently, MC and methanol are converted to DMC and ammonia (190 °C, 40 bar) over a ZnO-Al2O3 catalyst in a fixed-bed reactor. Methanol and DMC form an azeotrope; extractive distillation with methyl isobutyl ketone (MIBK) as entrainer is used to separate the azeotropic mixture. The reactor model for the reaction towards DMC based on kinetic rate expressions, showed that a long residence time (>10 h) and a relatively high MeOH:MC molar feed ratio of 6 are required to achieve reasonable single-pass conversions (15 %). This resulted however in an unrealistically large reactor volume and a large methanol load on the process. A feasibility study was done in order to improve the performance of the process. It was calculated that with a MeOH:MC ratio of 2 and a single-pass conversion of MC of 30 % the process would become technically feasible; the reactor volume decreased from 5,000 m3 to 600 m3 and the energy consumption of the process was decreased from 238 MW to 50 MW. A Pinch analysis showed that maximally 6 MW could be saved with heat integration, which corresponds to approximately 2 M$/y savings on energy costs. To produce 86 kt/y of DMC, the required amounts of raw materials are 80 kt/y of methanol and 58 kt/y of CO2, which results in an overall DMC yield from methanol of 38 %. The required total capital investment of the process is 110 M$. Economic feasibility depends on the DMC selling price. A price range between 800 and 1,100 $/t was assumed. For 800 $/t it is not possible to repay the capital investment within an assumed lifetime of 10 years and the process would therefore not be profitable. The break-even point is at 845 $/t. For a selling price of 1,100 $/t the gross profit becomes 22 M$/y, with a payback period of 3 years and a return on investment of 20 %.

Module layout optimization using a genetic algorithm in light water modular nuclear reactor power plants

Abstract 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.

Сравнительный анализ инвестиционной привлекательности энергоблоков АЭС на базе реакторных модулей малой и средней мощности и реактора большой мощности

Сравнительный анализ инвестиционной привлекательности энергоблоков АЭС на базе реакторных модулей малой и средней мощности и реактора большой мощности

Risk and regulatory considerations for small modular reactor emergency planning zones based on passive decontamination potential

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/MWth (3.1, 3.9, and 2.5 petabecquerels/MWth) 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.

Minor actinide balance reduction in ESFR-SMART

New safety measures were proposed recently for ESFR, a design for a large European sodium-cooled fast reactor. The fissile height was reduced by 25% and 5% in the inner and outer cores, respectively; a lower fertile blanket was implemented. A unique fissile enrichment was chosen. Additional fuel subassemblies (FAs), passively operating safety rods and corium discharge tubes were introduced. In the current core, the sodium void effect is strongly reduced, that is favourable for reactor safety, but the Am and Cm balances, i.e. their mass variations under irradiation, remain positive. In the paper we investigate options for Am incineration in the core radial and lower fertile blankets by introducing there a mixture of U and Am oxides instead of U oxide and/or steel. With this mixture in the radial blanket instead of steel, the Am and total minor actinide balances approach negative and zero values, respectively. With this mixture instead of U oxide and steel in the lower blankets, these balances are negative; the sodium void effect is lower. The modified cores produce more Pu and Cm.

Architectural model driven dependability analysis of computer based safety system in nuclear power plant

Abstract The most important non-functional requirements for dependability of any Embedded Real-Time Safety Systems are safety, availability and reliability requirements. System architecture plays the primary role in achieving these requirements. Compliance with these non-functional requirements should be ensured early in the development cycle with appropriate considerations during architectural design. In this paper, we present an application of system architecture modeling for quantitative assessment of system dependability. We use probabilistic model checker (PRISM), for dependability analysis of the DTMC model derived from system architecture model. In general, the model checking techniques do not scale well for analyzing large systems, because of prohibitively large state space. It limits the use of model checking techniques in analyzing the systems of practical interest. We propose abstraction based compositional analysis methodology to circumvent this limitation. The effectiveness of the proposed methodology has been demonstrated using the case study involving the dependability analysis of safety system of a large Pressurized Water Reactor (PWR).

Mixing Characteristics of Gas and Liquid Phases in Bubble Column Reactors from Virtual Tracer Simulation

A large body of literature is available in the field of bubble column reactors, focusing mainly on hydrodynamics. However, the mixing pattern, extent of back-mixing, and effect of scale on mixing are not well investigated, especially for large-size columns, partly because tracer experimental measurements in large-scale columns are difficult. This can be alleviated by computational experiments. This study focuses on using a virtual tracer experiment coupled with a two-phase three-dimensional computational fluid dynamics (CFD) simulation to investigate mixing characteristics in large reactors. The results are validated by comparison with experiments. Both gas- and liquid-phase mixing patterns are reported in terms of dispersion coefficient and mixing time, and information is provided for the diameter range 0.14–0.90 m and for various height/diameter ratios. Empirical correlations and the scale effect on dispersion coefficient are provided. The results are useful in the context of dispersion-based phenomenol...

CFD modeling a fluidized bed large scale reactor with various internal elements near the heated particles feeder

A numerical study of the fluidized bed industrial reactor in the presence of various internal elements is carried out by CFD methods A simple reactor heating efficiency function of chemical reaction with heat absorption is considered. The main emphasis is placed on the circulation flows of the catalyst particles and heating of the reactor. The analysis of the impact of various design elements on the heating efficiency of the reactor is carried out. Particular attention is paid to the possibility of baffles applying, which allows redirecting the flow. This effect may have an especially important value when the rapid heating of the reactor is required for temperature dependent reactions. The influence of heated catalyst feeder design on the efficiency of whole reactor heating is educed. The influence of the fractional composition of the catalyst, namely the presence of fine particles, on the reactor heating efficiency for different reactor design features is also studied. The results are carried out for a specific reactor example, but contribute to the overall branch of fluidized bed engineering.

The RAPID Fission Matrix Approach to Reactor Core Criticality Calculations

The Real-time Analysis for Particle transport and In-situ Detection (RAPID) code uses a unique, extremely fast, fission matrix–based methodology to compute the eigenvalue, and three-dimensional, pinwise fission source distribution for reactor, spent fuel pool, and spent fuel cask problems. In this paper, the RAPID fission matrix method is described and analyzed for application to several large pressurized water reactor problems, based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Monte Carlo Performance Benchmark problem. In the RAPID methodology, fission matrix coefficients precalculated using the Serpent Monte Carlo code, are then coupled together and solved for different core arrangements. A boundary correction method is used to obtain more accurate fission matrix values near the radial and axial reflectors. Eigenvalues and fission source distributions are compared between RAPID and Serpent reference calculations. In most cases, the eigenvalue differences between methods are less than 10 pcm. For a uniform core model, pinwise fission distributions between the methods differ by a root-mean-square value of , compared to a Serpent uncertainty of . The pinwise, axially dependent (100 axial levels) differences are , compared to a similar Serpent uncertainty of . To achieve these levels of uncertainty, the RAPID calculations are over 2500 times faster than Serpent, not counting the precalculation time.

Scalability of multitube membrane modules for hydrogen separation: Technical considerations, issues and solutions

Palladium membrane technology has shown promising features for the development of a sustainable hydrogen economy. Nonetheless, the contribution of a palladium membrane technology to economic and societal development requires its commercialization, diffusion and utilization. To generate enough incentives for commercialization, it is necessary to demonstrate the scalability and robustness of the membranes in industrial settings. Consequently, this work utilizes pilot-scale experimental data generated under industrial conditions to validate a Computational Fluid Dynamics (CFD) model, which was up-scaled and utilized to determine the intrinsic phenomena of palladium membrane scale up. This study reveals the technical/engineering requirements for the effective design of large scale multitube membrane modules. Mass transfer limitations and concentration polarization effects were studied quantitatively with the developed model using the defined parameters Concentration Polarization Coefficient (CPC) and Effective Average CPC (EAC). Methods for diminishing the concentration polarization effect were proposed and tested through the simulations such as i) increasing convective forces and ii) designing baffles to create gas recirculation. For scaled-up membrane modules, mass transfer limitation is an important parameter to consider as large modules showed severe concentration polarization effects. Certainly, this work shows for the first time the main features required when designing large scale membrane reactor modules.

Compressed Sensing Artificial Neural Network for Reactor Core Flux Mapping

This paper presents a novel technique of signal recovery from sparse measurements based on compressed sensing (CS) and artificial neural network (ANN) and its application to core flux mapping in Advanced Heavy Water Reactor (AHWR). In large size nuclear reactors, neutron flux distribution undergoes continuous variation due to routine perturbations, nonuniform burn up at different locations, xenon oscillation, etc. An online flux mapping system (FMS) is needed to continuously monitor neutron flux distribution and display to the operator. FMS employs a suitable algorithm to estimate the core flux distribution from the measurements of only a few in-core detectors. The proposed method called CS-ANN methodology first uses CS technique for flux mapping along the vertical direction at the detector housing locations (1-D) and subsequently ANN technique in several horizontal planes (2-D) for estimating the 3-D neutron flux profile under different operating conditions. Error in the estimation using the proposed CS-ANN methodology has been compared with other existing methods and found to be significantly lower.

Enhancement of sonochemical oxidation reactions using air sparging in a 36 kHz sonoreactor

The effect of air sparging on sonochemical oxidation reactions was investigated using a relatively large reactor equipped with a 36 kHz transducer module at the bottom. KI dosimetry and luminol techniques were used for quantitative and qualitative analysis of the reactions. The cavitation yield increased and then varied minimally as the liquid height increased from 1λ (42 mm) to 8λ (333 mm) with no air sparging. The flow rate of the air used for sparging and the position of the sparger significantly affected the extent of the sonochemical oxidation reactions. A significant enhancement in the sonochemical oxidation by air sparging was observed for higher liquid height and higher flow rate conditions at a constant input power. This enhancement is attributed to the violent mixing effect and the significant change in the sound field and cavitation-active zone in the liquid. Higher sonochemical activity was obtained when air sparging was applied closer to the transducer module at a higher flow rate. Imaging the motion of the liquid surface and sonochemiluminescence revealed that the instability of the liquid body was directly related to the sonochemical activity.

Simulations of SPERT-IV D12/15 transient experiments using the system code THERMO-T

Best-estimate codes, which couple neutronic and thermal-hydraulic solvers, are mainly used for safety analyses of nuclear power plants. During the past decade, the application of these codes to research reactors gained considerable interest and many improvements were presented to them. The increasing interest in the application of best-estimate codes to safety analyses of research reactor is largely driven by advancements in this field concerning power reactors and the diffusion of knowledge and capabilities to smaller, more diverse systems. The current study is a continuous effort in this framework and presents the coupled neutronic and thermal-hydraulic code development for the analysis of protected and unprotected transient behavior of research reactors. The coupling between neutronic and thermal-hydraulic processes is realized by considering the mutual feedbacks between them; the fuel and coolant properties (temperatures and density) variation affect the core's reactivity and hence the neutronic fission chain reaction, which in turn affects the fuel and coolant properties via a heat generation model for the reactor's power. More specifically, this study deals with the extension of the thermal-hydraulic model to the two-phase flow regime of the THERMO-T code. The extended THERMO-T model is validated against experimental results from the SPERT-IV, which was driven mainly by the coolant density reactivity feedback. This allows a more accurate evaluation of the adequacy of available and relevant two-phase flow models and correlations, which are selected from the domain of large power reactors. This is done in order to encourage and ensure standardization of modeling procedure of all types of reactors as part of the international community's continuous efforts towards this goal.

US nuclear power: The vanishing low-carbon wedge

Nuclear power holds the potential to make a significant contribution to decarbonizing the US energy system. Whether it could do so in its current form is a critical question: Existing large light water reactors in the United States are under economic pressure from low natural gas prices, and some have already closed. Moreover, because of their great cost and complexity, it appears most unlikely that any new large plants will be built over the next several decades. While advanced reactor designs are sometimes held up as a potential solution to nuclear power’s challenges, our assessment of the advanced fission enterprise suggests that no US design will be commercialized before midcentury. That leaves factory-manufactured, light water small modular reactors (SMRs) as the only option that might be deployed at significant scale in the climate-critical period of the next several decades. We have systematically investigated how a domestic market could develop to support that industry over the next several decades and, in the absence of a dramatic change in the policy environment, have been unable to make a convincing case. Achieving deep decarbonization of the energy system will require a portfolio of every available technology and strategy we can muster. It should be a source of profound concern for all who care about climate change that, for entirely predictable and resolvable reasons, the United States appears set to virtually lose nuclear power, and thus a wedge of reliable and low-carbon energy, over the next few decades.

Robust Optimal Integral Sliding Mode Controller for Total Power Control of Large PHWRs

This paper presents the design of an optimal integral sliding mode controller (OISMC) to control the total power of a large pressurized heavy water reactor (PHWR). Considering the dynamics of systems relevant to total power control loop of the large PHWR, the single input and single output model adopted in this paper is characterized by 21 state variables. The nonlinear model is linearized around steady-state full power operating point to obtain a linear model for which an OISMC is designed. Optimal controller which guarantees minimum control effort is combined with an ISMC to ensure the robustness of the system in the presence of uncertainties and external disturbances. Stability of the proposed controller is proven by means of Lyapunov approach. From nonlinear simulation studies, performance of the proposed OISMC is found to be satisfactory.

RETRACTED: Computational Fluid Dynamics Modeling of the Catalytic Partial Oxidation of Methane in Microchannel Reactors for Synthesis Gas Production

This paper addresses the issues related to the favorable operating conditions for the small-scale production of synthesis gas from the catalytic partial oxidation of methane over rhodium. Numerical simulations were performed by means of computational fluid dynamics to explore the key factors influencing the yield of synthesis gas. The effect of mixture composition, pressure, preheating temperature, and reactor dimension was evaluated to identify conditions that favor a high yield of synthesis gas. The relative importance of heterogeneous and homogenous reaction pathways in determining the distribution of reaction products was investigated. The results indicated that there is competition between the partial and total oxidation reactions occurring in the system, which is responsible for the distribution of reaction products. The contribution of heterogeneous and homogeneous reaction pathways depends upon process conditions. The temperature and pressure play an important role in determining the fuel conversion and the synthesis gas yield. Undesired homogeneous reactions are favored in large reactors, and at high temperatures and pressures, whereas desired heterogeneous reactions are favored in small reactors, and at low temperatures and pressures. At atmospheric pressure, the selectivity to synthesis gas is higher than 98% at preheating temperatures above 900 K when oxygen is used as the oxidant. At pressures below 1.0 MPa, alteration of the dimension in the range of 0.3 and 1.5 mm does not result in significant difference in reactor performance, if made at constant inlet flow velocities. Air shows great promise as the oxidant, especially at industrially relevant pressure 3.0 MPa, thereby effectively inhibiting the initiation of undesired homogeneous reactions.