Reactor Design Concepts Research Articles

Design and optimization of open-channel water ultraviolet disinfection reactor

Design and optimization have developed gradually from a methodology for academic attraction into a technology that has made a significant impact on the industrial growth. A careful and well-optimized design of open-channel water ultraviolet (UV) disinfection reactor ensures the achievement of minimum UV dose required, as well as allowing the minimum operational costs. Such UV reactors are mostly utilized in water treatment plants throughout the world to provide the system for disinfecting the drinking water. Yet, at present, lack of rigorous quantitative understanding of behavior in such reactor geometries is shown to limit the versatile and efficient optimization of UV reactor. A new UV reactor design concept is essential that should be able to consider the three significant parameters, namely: uniform UV light propagation within the open-channel UV reactor, providing optimum lamp positioning within the volume of the UV reactor and should allow optimum flow rate of the water. It is approved that computational fluid dynamics (CFD) simulation can be used as a quick and cheap means for design and optimization of water disinfection UV reactor. A novel design and optimization approach which combines CFD modeling, design of experiment (DOE), response surface method (RSM) and goal-driven optimization (GDO) was presented for open-channel water UV disinfection reactor. This methodology includes a heuristic approach from scratch to the final optimal solution. The optimal design variables which have the optimum value of the object functions were identified through proposed methodology. The results showed that the CFD method can be employed for estimation of the optimum design, even though the shape of the open-channel UV reactor is complex. The CFD results reveal that the UV dose distribution and flow of water were largely dependent on the lamp positioning. This paper demonstrates that UV reactor optimization in conceptual design is possible using a commercial CFD package. As a conclusion, the proposed design and optimization technique based on DOE, RSM and GDO could be a suitable technique prior to a final design of open-channel water UV disinfection reactor.

Small modular high temperature reactor optimisation – Part 1: A comparison between beryllium oxide and nuclear graphite in a small scale high temperature reactor

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.

Variable Reactivity Control in Small Modular High Temperature Reactors Using Moderation Manipulation Techniques

With extensive research being undertaken into small modular reactor design concepts, this has brought new challenges to the industry. One key challenge is to be able to compete with large scale nuclear power plants economically. In this article, a novel approach is applied to reduce the overall dependence on fixed burnable poisons during high reactivity periods within a high temperature graphite moderated reactor. To reduce the excess activity, we aim to harden the flux spectrum across the core by removing part of the central moderation column, thus breeding more plutonium, in a later period the flux spectrum is softened again to utilise this plutonium again. This provides a neutron storage effect within the 238U and the resulting breeding of Plutonium. Due to the small size and the annular design of the high temperature reactor, the central reflector is key to the thermalization process. By removing a large proportion of the central reflector, the fuel within the proximity of the central reflector are less likely to receive neutrons within the thermal energy range. In addition to this, the fuel at the extremities of the core have a higher chance of fission due to the higher number of neutrons reaching them. This works as a method of balancing the power distribution between the central and outside fuel pins. During points of low reactivity, such as the end of the fuel cycle, the central reflector can be reinserted and the additionally bred plutonium and U235 at the centre of the core will encounter a higher probability of fission due to more thermal neutrons within this region. By removing the central reflector, this provided a 320 pcm reactivity drop for the duration of the fuel cycle. The plutonium buildup provided additional fissile material up until the central reflector was reinserted. The described method created a two-fold benefit. The overall full power days within the core was increased by ~31 days due to the additional fissile material within the core and secondly the highest loaded power pins saw a 30% power reduction during the removal of the central reflector column.

Accident Process and Core Thermal Response During a Station Blackout Initiated Study for Small Modular Reactor

Interest in evaluation of reactor response and off-site consequences following beyond design basis accidents has significantly increased after Fukushima. Considering the inherent system safety of nuclear plants, experts begin to refocus on the development of small and medium nuclear reactors (SMRs). Based on proven and widely utilized traditional pressurized water reactor (PWR) technology, there are many kinds of small modular reactor design concepts, which are actively being developed in the USA, China, Japan, Russia, Korea, and other countries. And after Fukushima, interest in evaluation of reactor response and off-site consequences following beyond design basis accidents also has significantly increased. But due to the significant differences between the traditional distributed arrangement and integrated arrangement of SMRs, the variation trend of key parameters may be different during normal operation or accident process. So in this paper we simulated the small modular reactor severe accident by MELCOR. This papersummarizes the core thermal hydraulic response for a hypothetical severe accident caused by stationblackout at a small modular reactor using MELCOR. Analytical results for temperature distribution of the fuelpellets, the fuel cladding, flow rate of the coolant, andhydrogen mass changing over time arepresented.The analyses are focused on safety assessment of the reactor core for severe accidents and arepart of theoverall evaluation of safety features of the small modular reactor for residual risk posed by severeaccidents.

Core design of PWR-type seed-blanket core breeder reactor with tightly packed fuel assembly

Pressurized water reactor (PWR) is the reactor type with the most abundant operation experience in the world. However, studies on designing PWR-type fast reactors have been limited and there have not been any PWR-type fast breeder reactor design concepts. In this study, the concept of seed-blanket PWR-type breeder reactor with tightly packed fuel assembly (TPFA) has been developed by coupled three dimensional neutronics and thermal-hydraulics core calculations. For the seed-blanket heterogeneous core using mixed oxide (MOX) fuel and depleted uranium, it has been shown that the core height is limited to about 1.0 m or less, in order to satisfy the design criterion of negative void reactivity. Moreover, it has been shown that increasing the core power density is difficult, as it leads to substantial increase in the core pressure drop. Consequently, the core characteristics are featured by low breeding performance with compound system doubling time (CSDT) of about 150 years or more (fissile plutonium surviving ratio of below 1.01) and low thermal power of about 700 MW or less. For the core using 4.95 wt% enriched uranium for the blanket assembly, it is possible to improve the void reactivity characteristics, breeding performance and thermal power by reducing reactivity difference between the seed and the blanket fuel assemblies. By utilizing enriched uranium, the concept of breeding PWR core with CSDT of 60 years and thermal power of 1000 MW has been shown. In addition, PWR-type breeder reactor concept using enriched uranium that the fissile surviving ratio including uranium and plutonium exceeds 1 was shown for the first time.

Preliminary design of the I2S-LWR containment system

The Integral Inherently Safe Light Water Reactor (I2S-LWR) is a novel reactor design concept which aims at delivering an electric power output level comparable to that of large LWRs (approximately 1000 MWe), while at the same time achieving an overall level of safety that is enhanced with respect to large Generation III+ LWRs. One of the main safety goals is to achieve indefinite cooling following design basis accidents (DBAs) using atmosphere air as ultimate heat sink. In order to accommodate these goals, the I2S-LWR incorporates several innovative safety features, including an integral Reactor Pressure Vessel (RPV), enhanced passive Decay Heat Removal (DHR) systems and several containment passive cooling systems. The present work is focused on a passive and reliable containment design, which plays a significant role in LOCA scenario and is the last boundary to prevent the release of radioactivity to the environment. In this paper, several innovative passive systems located in the I2S-LWR containment are proposed, including Core Make-up Tank (CMT), Accumulator (ACC), Passive Suppression System (PSS), Automatic Depressurization System (ADS), Passive Containment Cooling System (PCCS) and Passive Reactor Cavity Cooling System (PRCCS). The best-estimate thermal-hydraulic code RELAP5 has been used to model the I2S-LWR RPV and containment passive systems. The inadvertent opening of PORV (Power Operated Relief Valve) accident scenario has been simulated in order to study the containment response, including the coupling between RPV and containment. The results show that, through the activation of the ADS, pressure equilibrium between containment and reactor pressure vessel is achieved, while maintaining the reactor core covered at all times. The containment passive cooling systems ensure that the containment pressure remains at acceptable levels throughout the transient. A sensitivity study on the number of operating ADS valves, and on the availability of accumulators is also performed.

Stress corrosion cracking susceptibility of thermally-aged Type 310S stainless steels in supercritical water

The effect of thermal ageing on the stress corrosion cracking (SCC) susceptibility of Type 310S stainless steel in supercritical water (coolant for a supercritical water cooled reactor design concept) was evaluated. SCC was evaluated using the slow strain rate testing technique, straining samples to 10% in a flowing autoclave test facility using 25 MPa SCW at 500 °C with 20 ppm dissolved oxygen. The localized deformation model for SCC initiation and a void-assisted film-cleavage model for SCC propagation can be used to explain the SCC susceptibility exhibited by the thermally-aged Type 310S stainless steel in supercritical water.

Effect of Thermal Aging on the Intergranular Stress Corrosion Cracking Susceptibility of Type 310S Stainless Steel

The effect of thermal aging on the intergranular stress corrosion cracking (IGSCC) susceptibility of Type 310S stainless steel was investigated. Samples were solution-annealed (baseline), sensitized (short-term aging), and thermally-treated (long-term aging) to form Cr-rich M23C6 carbides and sigma (σ) phase precipitate particles, respectively, on the grain boundaries. The thermal-aging heat treatments represent two limiting exposure conditions expected in-service for a fuel cladding in the Canadian Generation IV (supercritical water-cooled reactor) design concept and were considered to better understand the risk factors associated with an in-service IGSCC damage mode. The sensitized treatment exhibited the highest relative degree of sensitization (measured by double loop electrochemical potentiokinetic reactivation testing in 2 M H2SO4 + 0.01 M KSCN) and IGSCC susceptibility (measured by slow strain rate testing [SSRT] in a hot alkaline solution). The relative susceptibility reflects the controlling role played by Cr-depleted zones, which were only observed in the sensitized (short-term aging) material. The relatively large σ phase grain boundary particles in the thermally-treated (long-term aging) material preferentially cracked during SSRT, suggesting an apparent IGSCC susceptibility at the sample surface. However, the absence of Cr-depleted zones in this material prevented IGSCC from occurring, despite preferential cracking in the σ phase grain boundary particles. The dominant fracture mode in the hot alkaline solution was transgranular mixed-mode cracking. The implications of the results are discussed within the context of risk factors associated with an in-service IGSCC susceptibility in supercritical water.

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

A Multi-Application Small Light Water Reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s ability to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with computational fluid dynamics simulation for confirmation. The corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.

Reactor Performance and Design Concept in Additively Manufactured Milli-Scale Reactors

Millireactor or minireactor technology in small scale production has gained increasing attention in microreactor technology research because high pressure drops, unequal flow distribution into micrometer scale channels, and channel blocking can be reduced in larger scale. In this work, additive manufacturing suitability for reactors has been studied based on reactor surface properties. Mixing and heat transfer have vital role in reactor design when exothermic and fast or relatively fast reactions are considered. Here, mixing efficiencies of Hartridge-Roughton reactor and its modifications have been evaluated. The surface roughness and surface rugosity had influence in experimentally determined mixing efficiencies which made a clear difference when comparing simulation results. Reactor temperature fields were experimentally measured with IR camera technology as well as the heat flows were simulated for instantaneous acid-base neutralization reaction. The milli-scale mixer design cycle consisting Computational Fluid dynamics (CFD), additive manufacturing and experimentation is proposed. The proposed methodology is geometry independent and millireactor design steps here are discussed in detail. The design tasks are introduced for illustrating the proposed design concept for reactor geometry optimization.

Natural circulation thermal-hydraulics model and analyses of “SLIMM” – A small modular reactor

The Scalable LIquid Metal cooled small Modular (SLIMM) reactor design concept has recently been developed at the University of New Mexico’s Institute for Space and Nuclear Power Studies. It generates 10–100MWth for ∼66 and 5.9 full power years, respectively, without refueling. Natural circulation of in-vessel liquid sodium cools the reactor core during nominal steady-state operation and after shutdown, with the aid of an in-vessel chimney (2–8m tall) and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. A natural circulation thermal-hydraulics model of the SLIMM reactor is developed to investigate the effects of thermal power, in-vessel chimney height, and HEX design on the circulation rate and core exit temperature of in-vessel liquid sodium and the pressure losses during steady-state operation. The HEX design parameters investigated are the number, pitch and tube diameter of the helical coils, and the temperature pinch between the in-vessel sodium flow in the downcomer and the secondary sodium flow through the HEX coils. Increasing the reactor thermal power or the in-vessel chimney height increases the flow rate of the in-vessel liquid sodium by natural circulation, but decreases its exit temperature from the reactor core. Conversely, increasing the number or the tube diameter of the HEX helical coils decreases the flow rate of in-vessel liquid and increases its exit temperature from the reactor core, and decreases the HEX total height. Depending on the steady-state reactor thermal power and HEX design, the in-vessel sodium flow through the reactor core contributes ∼64–69% of the total pressure losses, while the sodium flow in the HEX contributes ∼32–29%, respectively

Electrochemical jet-cell for the in-situ generation of hydrogen peroxide

In this work, a proof of concept of a novel reactor design for the in-situ electrochemical production of H2O2 from oxygen reduction reaction is presented for the first time. The innovative design incorporates a venturi-based jet aerator to supply atmospheric oxygen without additional energy consumption to a 3D flow-through modified carbon felt (CF) cathode. Preliminary experiments confirmed that electro-generation of hydrogen peroxide is possible in a system as the one proposed. Comparison with a flow-by cell with a gas diffusion cathode under similar conditions revealed that current efficiency towards hydrogen peroxide accumulation is even higher (72 vs 65% at 1h) than in the case of the conventional system. Jet aerator stands as a promising oxygen supply thanks to its excellent performance and both low investment cost and energy consumption. Considering all the above, the electrochemical jet cell stands as a rather promising design for the efficient hydrogen peroxide electro-generation.

Subchannel analysis of Al2O3 nanofluid as a coolant in VMHWR

AbstractThe main objective of this study is to predict the thermal hydraulic behavior of nanofluids as the coolant in the fuel assembly of variable moderation high performance light water reactor (VMHWR). VMHWR is the new version of high performance light water reactor (HPLWR) conceptual design. Light water reactors at supercritical pressure (VMHWR, HPLWR), being currently under design, are the new generation of nuclear reactors. Water-based nanofluids containing various volume fractions of Al2O3 nanoparticles are analyzed. The conservation equations and conduction heat transfer equation for fuel and clad have been derived and discretized by the finite volume method. The transfer of mass, momentum and energy between adjacent subchannels are split into diversion crossflow and turbulent mixing components. The governed non linear algebraic equations are solved by using analytical iteration methods. Finally the nanofluid analysis results are compared with the pure water results.

Thermal hydraulics analysis of the Advanced High Temperature Reactor

The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

Improved delayed detached eddy simulation of a randomly stacked nuclear pebble bed

The high temperature reactor (HTR) design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in the event of an accident such as loss of coolant. However, the possible appearance of hot spots in a pebble bed core design may affect the integrity of the pebbles and the fuel. This has drawn the attention of several scientists to understand this highly three-dimensional complex phenomenon. To obtain accurate predictions based on techniques such as DNS and LES, for a realistic (even for a small size) pebble bed flow, is still computationally too expensive and not foreseeable in the near future. On the other hand, the prediction capabilities of turbulence modelling approaches such as RANS and hybrid RANS-LES methods for such complex flow regime have not yet been rigorously evaluated. In this paper, numerical simulations of a limited sized randomly stacked bed of spherical pebbles are performed by using improved delayed detached eddy simulation (IDDES). The pebble bed configuration analysed consists of approximately 30 pebbles, which are randomly stacked to represent the core of an HTR. The obtained results are compared (qualitatively and quantitatively) with the available reference LES for validation purposes. The results show that the selected IDDES based on the SST k-ω model is able to reproduce the overall flow topology. The mean flow and thermal fields are found to be in good agreement with the LES. However, the second order statistics have shown significant disagreement. Nevertheless, these results are explicitly discussed in detail to understand the flow and thermal fields appearing in this complex flow configuration.

Feasibility assessment of the once-through thorium fuel cycle for the PTVM LWR concept

This paper investigates the feasibility of a once-through thorium fuel cycle for the novel reactor-design concept named the pressure tube light water reactor with variable moderator control (PTVM LWR). The PTVM LWR operates in a “breed & burn” mode, which makes it an attractive system for utilizing thorium fuel in a once-through mode. The “breed & burn” mode can emphasize the in situ generation as well as incineration of 233U, which are the basic foundations of the once-through thorium fuel cycle. The PTVM LWR concept makes use of a seed–blanket geometry, whereby the core is divided into separated regions of thorium-based fuel channel assemblies (blanket) and low-enriched uranium (LEU) based fuel channel assemblies (seed). A novel fuel in-core management scheme based on two separate fuel flow routes (i.e., seed route and blanket route) is proposed and analyzed. Neutronic performance analysis indicates that the proposed novel fuel in-core management scheme has the potential to utilize both LEU- and thorium-based fuel in an efficient manner. The once-through thorium cycle, presented and discussed in this paper, provide interesting research leads and can serve as a bridge between current LEU-based fuel cycles and a thorium fuel cycle based on recycling of 233U.

A Conceptual Reactor Design Based on Combination of Pressure Leaching - Electrochemical Deposition Techniques for Study of Gold Recovery from e-Waste

In Malaysia, the amount of electronic waste (e-waste) generated in 2008 was stated to be around 688,000 metric tonnes and was forecasted to increase to 1.11 million metric tonnes in the year 2020 [1]. Typical e-waste contains some elements of precious metal, which is mainly generated from the result of the revolution of computer technology. Currently, the most common technique being practiced to recover gold from e-waste is through cyanide leaching system and subsequent gold recovery techniques which may include electrodeposition. However, due to toxicity of the harmful cyanide, a non-cyanide extraction route is now becoming a necessity to minimize safety and environmental issues. In order to maximize the efficiency and safety of gold recovery from e-waste, a non-cyanide route using a reactor based on electrodeposition technique has been proposed. This paper explains the conceptual reactor design which combines the pressure leaching process of gold from e-waste and the gold recovery using electrodeposition via a non-cyanide route as a safer and environmental friendly process in recovering gold from e-waste.

석탄의 고품위화를 위한 기류건조기와 로터리킬른 건조기의 성능 비교평가

Drying, which is the oldest and most energy-intensive process, is an essential process for treatment of solid product. The specific procedure to design and evaluate the dryers, which are a rotary kiln dryer and a flash dryer, in case of drying the high-moisture coal was described. From determination of size to the heat and mass balance in one-dimensional model were conducted to evaluate the performance of dryers. Heat consumption, inlet gas temperature and size of the dryers were compared between a rotary kiln dryer and a flash dryer. Further considerations to evaluate the reactor elaborately were also discussed. Performance simulation of dryers along with the design procedure described here will provide helpful basis for understanding the concept of reactor design.

Robust autothermal microchannel reactors

Autothermal microchannel reactors are intensified process units that bring significant energy efficiency benefits over their conventional counterparts. Efficiency gains are obtained, however, at the cost of operational challenges. These stem from the loss of control handles that is inherent to combining several unit operations in a single physical device. In this paper, we investigate the impact of two recently proposed reactor design concepts (a segmented catalyst macromorphology and an embedded layer of phase change material) aimed at improving the steady state energy distribution and, respectively, preventing the advent of hotspots during transient operation, on reactor dynamics and control. Using an autothermal microchannel reactor coupling steam methane reforming with methane catalytic combustion as a prototype system, we demonstrate through rigorous simulations that these design innovations have a synergistic effect, resulting in superior steady-state performance and excellent disturbance rejection ability.

Creep deformation and rupture behavior of CLAM steel at 823 K and 873 K

China Low Activation Martensitic (CLAM) steel is selected as the candidate structural material in Fusion Design Study (FDS) series fusion reactor conceptual designs. The creep property of CLAM steel has been studied in this paper. Creep tests have been carried out at 823K and 873K over a stress range of 150–230MPa. The creep curves showed three creep regimes, primary creep, steady-state creep and tertiary creep. The relationship between minimum creep rate (ε̇min) and the applied stress (σ) could be described by Norton power law, and the stress exponent n was decreased with the increase of the creep temperature. The creep mechanism was analyzed with the fractographes of the rupture specimens which were examined by scanning electron microscopy (SEM). The coarsening of precipitates observed with transmission electron microscope (TEM) indicated the microstructural degradation after creep test.