Next Generation Nuclear Research Articles

The use of burnable absorbers integrated into TRISO/QUADRISO particles as a reactivity control method in a pebble-bed HTR reactor fuelled with (Th,233U)O2

High-temperature gas-cooled reactors (HTGRs) are one of the promising candidates for the next generation of nuclear reactors due to their inherently and passively safe in normal operation and even under accident conditions. There are two HTGR concepts, one based on a pebble-bed core, and the other is a prismatic core. One of the key differences between the pebble-bed core and the prismatic core is the complexity of applying the seed-blanket (SB) concept to the internal production of U-233 from Th-232 in order to extend the cycle length of the reactor. Actually, an enhancement of the operating cycle length in the pebble-bed reactors fuelled with (Th,233U)O2 can be reached by using high U-233 ratios, but these ratios could result in high excess reactivity at the beginning of life (BOL). In order to suppress the excess reactivity phenomenon at the early stage of the fuel cycle, to flatten the reactivity curve as a function of time, a new reactivity control method has been investigated by incorporating burnable absorbers into the fuel particles. This purpose is set up by introducing two varieties of BAs, including Gadolinia and Erbia, either as a homogeneous mixture with the fuel kernel in the TRISO particles or as an additional layer next to the fuel kernel producing the quadruple isotropic (QUADRISO) particles. The present work aims at characterizing the effect of these different loading patterns on some neutronic parameters that define the reactor core under BOL conditions as well as during burnup. In this attempt, the nominal UO2 (17 wt% U-235) fuelled core was used as a reference, while the other models adopt (Th,233U)O2 (17 wt% U-233) fuel with Gadolinia and/or Erbia with different loading patterns. The calculation results show that excess reactivity can be compensated during the early stages of core life and flattened during the cycle by adding Erbia doped with a few ppm of Gadolinia to the TRISO/QUADRISO particles. In these updated configurations, the new core life achieved is approximately 678 Effective Full Power Days (EFPDs), equivalent to 52.5 GWd/t on a single-batch discharge burnup. This means that from the core fuelled with UO2, which has a core life of 537 EFPDs (37 GWd/ton), an average period of 141 EFPDs is extended.

3D Printing to Increase the Flexibility of the Chemical Synthesis of Biologically Active Molecules: Design of On-Demand Gas Generation Reactors.

The development of new drugs is accelerated by rapid access to functionalized and D-labeled molecules with improved activity and pharmacokinetic profiles. Diverse synthetic procedures often involve the usage of gaseous reagents, which can be a difficult task due to the requirement of a dedicated laboratory setup. Here, we developed a special reactor for the on-demand production of gases actively utilized in organic synthesis (C2H2, H2, C2D2, D2, and CO2) that completely eliminates the need for high-pressure equipment and allows for integrating gas generation into advanced laboratory practice. The reactor was developed by computer-aided design and manufactured using a conventional 3D printer with polypropylene and nylon filled with carbon fibers as materials. The implementation of the reactor was demonstrated in representative reactions with acetylene, such as atom-economic nucleophilic addition (conversions of 19–99%) and nickel-catalyzed S-functionalization (yields 74–99%). One of the most important advantages of the reactor is the ability to generate deuterated acetylene (C2D2) and deuterium gas (D2), which was used for highly significant, atom-economic and cost-efficient deuterium labeling of S,O-vinyl derivatives (yield 68–94%). Successful examples of their use in organic synthesis are provided to synthesize building blocks of heteroatom-functionalized and D-labeled biologically active organic molecules.

Schemes for producing multi-energy fast neutrons using a low-energy ion accelerator

In this article we explore neutron producing schemes using a compact accelerator for producing fast neutrons for nuclear data measurements relevant to GEN-IV reactors and Accelerator Driven Systems and also for other applications such as radioisotopes production. The compact accelerator which consists of an Electron Cyclotron Resonance ion source along with a Radio Frequency Quadrupole and Drift Tube Linac is expected to provide p,d and 7Li beams of energies in the range of 0.5–3 MeV/u. Two different neutron producing schemes are examined for their suitability for the above mentioned applications. In the first scheme, using the 7Li(d,n)2 4He reaction, a variety of neutron energies ranging from 1 MeV to 16 MeV with reasonable flux and energy-time characteristics can be generated making it suitable for studies in (n,cp), (n,f) and (n,Xn) reactions important for nuclear fusion technology, security applications and radioisotope production. In the second scheme, the p(7Li,n)7Be reaction can be used to produce quasi mono-energetic neutrons in inverse kinematics mode with very little background making it suitable for high precision nuclear data measurements relevant to GEN-IV reactors and Accelerator Driven Systems.

A Finite Element Penalized Direct Forcing Immersed Boundary Method for infinitely thin obstacles in a dilatable flow

In the framework of the development of new passive safety systems for the second and third generations of nuclear reactors, the numerical simulations, involving complex turbulent two-phase flows around thin or massive inflow obstacles, are privileged tools to model, optimize and assess new design shapes. In order to match industrial demands, computational fluid dynamics tools must be the fastest, most accurate and most robust possible. To face this issue, we have chosen to solve the Navier-Stokes equations using a projection scheme for a mixture fluid coupled with an Immersed Boundary (IB) approach: the penalized direct forcing method – a technique whose characteristics inherit from both penalty and immersed boundary methods – adapted to infinitely thin obstacles and to a Finite Element (FE) formulation. Various IB conditions (slip, no-slip or Neumann) for the velocity on the IB can be managed by imposing Dirichlet values in the vicinity of the thin obstacles. To deal with these imposed Dirichlet velocities, we investigated two variants: one in which we use the obstacle velocity and another one in which we use linear interpolations based on discrete geometrical properties of the IB (barycenters and normal vectors) and the FE basis functions. This last variant is motivated by an increase of the accuracy/computation time ratio for coarse meshes. As a first step, concerning academic test cases for one-phase dilatable-fluid laminar flows, the results obtained via those two variants are in good agreement with analytical and experimental data. Moreover, when compared to each other, the linear interpolation variant increases the spatial order of convergence as expected. An industrial test case illustrates the advantages and drawbacks of this approach. In a shortcoming second step, to face two-phase turbulent fluid simulations, some methodology modifications will be considered such as adapting the projection scheme to low-compressible fluid and immersed wall-law boundary conditions.

ПЕРСПЕКТИВЫ АТОМНОЙ ЭНЕРГЕТИКИ В ОБЕСПЕЧЕНИИ ЭНЕРГЕТИЧЕСКОЙ И ЭКОЛОГИЧЕСКОЙ БЕЗОПАСНОСТИ РОССИИ

Large-scale nuclear energetics can satisfy demands for all kinds of energy, i.e. it can secure energy safety of Russia and the whole humankind; however, this is associated with a number of daunting problems. With that, this approach is a priority for Russia. The State Corporation RosAtom is involved in the development of nuclear reactors in Russia and abroad on the conditions that the reactors will be supplied with nuclear fuel from Russia and the spent fuel will be returned to Russia for conversion into mixed uranium and plutonium oxide (MOX) fuel. In the city Zheleznogorsk near Krasnoyarsk, the first production line of a plant for treating 2000 tons of spent nuclear fuel annually has been already launched. The principal strategic plan of RosAtom, which has been being realized currently, is to develop nuclear power production based on fuel recycling using fast neutron reactors for generation of plutonium, which may be used in nuclear weapons and is most hazardous for the biosphere. The possibility of accidents associated with radioactive discharges cannot be excluded, and the hazardousness of such accidents in increased by using plutonium-based fuels. The nuclear power-based approach to energy production is costly but also dangerous not only for Russia.

Isolation of a heterotrophic nitrification-aerobic denitrification strain and identification of its potential electricity generation ability in microbial fuel cells

ABSTRACT To construct a simultaneous nitrification and denitrification-microbial fuel cell (SND-MFC) reactor for stable electricity generation, a heterotrophic nitrification-aerobic denitrification strain was isolated and purified from aerobic activated sludge from a sewage treatment plant. The strain with an optimal nitrogen removal performance, which was identified as Pseudomonas, was inoculated into the SND-MFC as a single strain. Different electrode materials and electrode distances (EDs) were investigated. The results showed that a maximum ammonium removal rate of 92.31% and a maximum power density of 134.28 mW/m3 were obtained by the SND-MFC using graphene film as the electrodes material. Decreasing the ED did not significantly improved the power generation performance of the pure strain SND-MFC at the initial stage. When the electrode distance of the SND-MFC was 4 cm, the best generation efficiency was achieved with a maximum power density of 151.84 mW/m3.

New Reactor Safety Measures for the European Sodium Fast Reactor—Part I: Conceptual Design

Abstract Following up the previous collaborative project on European sodium fast reactor (CP-ESFR project), the European sodium fast reactor—safety measures assessment and research tools (ESFR-SMART project) considers the safety objectives envisaged for generation-IV reactors, taking into account the lessons learned from the Fukushima accident, in order to increase the ESFR safety level. In accordance with these objectives, guidelines have been defined to drive the ESFR-SMART developments, mainly simplifying the design and using all the positive features of sodium fast reactors (SFR), such as low coolant pressure, efficiency of natural convection, possibility of decay heat removal (DHR) by atmospheric air, high thermal inertia, and long grace period before a human intervention is needed. In this paper, a set of new ambitious safety measures is introduced for further evaluation within the project. The proposed set aims at consistency with the main lines of safety evolutions since the Fukushima accident. The paper gives a first review of the new propositions to enhance the ESFR safety, leading to a simplified reactor, forgiving and including a lot of passivity. This first version is supported by the various project tasks in order to assess the relevance of the whole design in comparison to the final safety objectives.

New Reactor Safety Measures for the European Sodium Fast Reactor-Part II: Preliminary Assessment.

The European Sodium Fast Reactor Safety Measures Assessment and Research Tools (ESFR-SMART) project offers innovative options for a sodium fast reactor (SFR) to improve its safety. This paper explains the preliminary calculations made on the main options to roughly verify their feasibility. Design propositions and calculations are here provided of the following innovative options: elimination of the safety vessel, innovative decay heat removal systems (DHRS), core catcher, thermal pumps, and secondary loops. In conclusion, all these options seem able to fulfill the key points of new safety rules for Generation-IV reactors. A status of the research and development (R&D) effort necessary to validate these new options is also proposed.

Structure and Heat Transfer in Zircaloy-4 Treated at High Temperatures.

Zircaloy-4 has an important role in the construction of generation III nuclear reactors. An important application is the fuel element sheath, which must have excellent corrosion resistance in the working environment, adequate mechanical characteristics and very good heat transfer properties from the combustible element to the coolant. The corrosion processes at high temperatures, the accidents that lead to significant increases in temperature and the structural transformations associated with them affect the heat transfer process. The paper presents research on the influence of high temperatures on the microstructure and thermal diffusivity of the zy-4 alloy. The samples were treated in air, at temperatures between 850 and 1050 °C for 60 min. The corrosion layers were characterized microstructurally and chemically. Furthermore, the transformations produced in the base material under the corrosion layer were analyzed. The values of thermal diffusivity were determined and correlated with the structural transformations. Considering the state of research on the materials appropriate to be used for new generation reactors, the current importance of third-generation reactors for energy systems and the fact that they will operate in the coming years, we consider that the study offers useful outcomes in the field of nuclear energy.

Nuclear Micro Power Reactor: The New Generation of Innovative Small Reactors

This short communication gives an overall account of small modular reactors and then walks through the nuclear micro reactors as the next generation of small modular reactors, which is the next wave of innovation for these SMRs. These next wave ride on the fact that future nuclear reactors are getting smaller and modular as well transportable. In this paper we are covering a summary and overall aspect of Generation IV (GEN-IV) or they are also known as Small Modular Reactors (SMRs) as well. In this book, we also, cover Nuclear Micro Reactor and its need and implementation within Department of Defense (DOD) military organizations.

Source term analysis and sorting scheme in post-irradiation spherical pyrometric elements

The 10 MW high-temperature gas-cooled reactor (HTR-10) is the most advanced of the fourth generation reactors, and has become a good test platform for very high-temperature operations. Spherical fuel elements are utilized in HTR-10, and the maximum temperature of the fuel element is one of the primary technical factors restricting the outlet temperature of the core. In our experiment, a pyrometric element with different wire melting points, is added to the HTR-10 reactor core to measure the maximum temperature distribution. How to sort out the pyrometric element, and analyze the source terms in post-irradiation spherical pyrometric elements when cycled out of the core, are the key issues of this study. This study can provide technical support for the internal temperature measurement, and analysis of the core, as well as for very high-temperature operation of HTR-10.

System analysis methodology for decision making in the design problems of new generation nuclear reactors

The main goal of the article is to summarize the results of many years of work on the creation of a unified methodology for designing new generation nuclear reactors that meet the requirements for generating electricity on a given scale, economic efficiency, inherent safety, and fuel supply. The methodology is based on the methods of systems analysis and operations research. Accidents with unacceptable releases of radioactive substances outside the NPP in new generation reactors can be eliminated by using core materials with specified (obtained within the same methodology) properties. The research methods are based on solving mathematical programming problems in different formulations, solving a number of multicriteria problems, using the non-formalized theory of conflict and informal procedures. As a result of the research, decision-making algorithms have been proposed; software has been developed and modernized to solve the problem of designing nuclear reactors. The practical significance of research in substantiating the fundamental possibility of creating safe nuclear power based on high-power reactors with lead cooling, cermet fuel and innovative structural materials.

Multi-Physics Model Development for Polonium Transport Behavior in a Lead-Cooled Fast Reactor

210Po, a highly toxic element with strong volatility, is one of the main source terms of a Gen-IV lead-cooled fast reactor (LFR). Therefore, the radioactive safety caused by 210Po has become an important topic in LFR-related research. In order to simulate the behavior of 210Po in an LFR, this work developed a multi-physics model of an LFR from the perspective of radioactive transport. Considering the effects of nuclide decay, cover gas leakage, containment ventilation, and Po aerosol deposition, a comprehensive simulation was carried out to evaluate the sensitivity of those effects on the 210Po distribution in detail. Preliminary results indicate that during normal operation, most of the 210Po in the LBE exist in the form of PbPo, and around 10–9 of 210Po could evaporate from the LBE into the cover gas, and then further leak into the containment. In addition, even if the leakage rate of 210Po in the cover gas into the containment is maintained at 5‰ per day, due to the deposition of Po aerosol, the 210Po contamination on the inner surface of the containment is still below the radioactivity concentration limits.

High performance solar receiver–reactor for hydrogen generation

This paper reports on the numerical analysis of a volumetric solar receiver-reactor for hydrogen production, using the 2-step reduction–oxidation cycle. A detailed parametric sweep covering hundreds of various parameter combinations is performed for a large solar reactor, using a transient physical model. We generate performance maps which are currently cost prohibitive via experimental or high–fidelity simulation studies. The following performance metrics are evaluated: solar to fuel efficiency, hydrogen yield, conversion extent and specific hydrogen yield. We show that the relations between the different performance metrics are complex, leading to different optimal points depending on the metric pursued. The daily hydrogen yield for a single reactor varied between 0.89 kg for an absorber thickness of 30 mm, and up to 1.04 kg for a 60 mm thick receiver, with solar to fuel efficiency values of 3.84% and 3.81% respectively. For a case with 45 mm thick receiver, an intermediate hydrogen yield of 0.94 kg is calculated, while exhibiting the highest efficiency (4.05%). The efficiency can be further increased to 5.86% by using a simple heat recovery system, and reach an upper limit of 21.16% with a more sophisticated heat recovery method.

Potential advantages of molten salt reactor for merchant ship propulsion

Operating costs of merchant ships, related to fuel costs, has led the naval industry to search alternatives to the current technologies of propulsion power. A possibility is to employ nuclear reactors like the Russian KLT-40S, which is a pressurized water reactor (PWR) and has experience on civilian surface vessels. However, space and weight are critical factors in a nuclear propulsion project, in addition to operational safety and costs. This work aims at comparing molten salt reactors (MSR) with PWR for merchant ship propulsion. The present study develops a qualitative analysis on weight, volume, overnight costs, fuel costs and nuclear safety. This work compares the architecture and operational conditions of these two types of reactors. The result is that MSR may produce lower amounts of high-activity nuclear tailings and, if it adopts the 233U-thorium cycle, it may have lower risks of proliferating nuclear weapons. Besides proliferation issues, this 4th generation reactor may have lower weight, occupy less space, and achieve the same levels of safety with less investment. Thus, molten salt regenerative reactors using the 233U-thorium cycle are potential candidates for use in ship propulsion.

Neutronic analysis of the ALLEGRO fast reactor core with deterministic ERANOS code and Monte Carlo Serpent code

The Gas-cooled Fast Reactor is an advanced concept selected to be part of the fourth generation of nuclear reactors. Before its development, the construction of the experimental demonstration reactor ALLEGRO is planned. Given the current interest in this innovative system, a neutronic study of the ALLEGRO reactor core is performed using the Monte Carlo Serpent calculation code and the deterministic ERANOS code. A detailed description of the core design and ceramic fuel composition is provided. To reduce the computational cost of the simulations for future applications, different cases are analyzed using different computational options in the TGV/VARIANT module and a 7-group energy structure. To validate the ERANOS models, the results are compared with the three-dimensional heterogeneous reference model developed in Serpent. The main parameters of the core, at the beginning of life, are calculated, such as the k-eff value, flux and power distributions, Doppler constant, effect of helium density on reactivity and the β-eff value. The main discrepancies in the results correspond to the diffusion and simplified transport calculations, due to the low density of helium. By using the 7-group structure, it was possible to reduce the calculation time with a reduced penalty in the precision of the results. Furthermore, the best agreement was obtained, with respect to the Serpent model, for the transport calculation (P3) with the 7-energy group structure. The k-eff and fuel isotope mass evolution during burnup is analyzed assuming an operating time of 365 days. The relative fuel fraction at the beginning and end of cycle, the breeding ratio and the average burnup are also given.

The state of the art of nuclear energy and its bibliometric analysis

Nuclear power plants, which started to be used in the military field in the 19th century, began to be used in electricity generation due to the oil crisis and increasing electricity need. It has many advantages as well as some disadvantages. Commonly used fossil, hydraulic, wind and solar energy systems likewise have both advantages and disadvantages. The emission of nuclear radiation is the most important risk. If there is no radiation emission, it will cause a less bad impact even if an accident happens. Also, waste management will be easier. With the development of technology, the production of new generation reactors can ensure that some disadvantages are eliminated or minimized. When looking at the studies carried out in WOS, studies on nuclear energy are related to environmental issues (waste management and occupational health). However, more studies should be carried out on two problems (waste management and occupational health) that are important in the operation of nuclear power plants.

Measurements by x-ray diffraction of the temperature dependence of lattice parameter and crystallite size for isostatically-pressed graphite

Synthetic polygranular graphites of various grades and manufacturing routes are used in nuclear reactors for power generation, and may be used in potential fourth generation and other advanced reactor designs that will operate at higher temperature. Attention is given in this paper to isostatically-moulded synthetic polygranular graphites with porosities in the range 8% to 18%. The lattice parameters a and c for the hexagonal graphite have been measured over the temperature range from room temperature to 800°C by x-ray diffraction. The variation with temperature of the crystal lattice parameters, coherence length (crystallite size) and microstrain are discussed with reference to the microstructure and the relative strength of the bonds in-plane and normal to the graphene layers.

Ultra-fine-grained and gradient FeCrAl alloys with outstanding work hardening capability

FeCrAl alloy is a promising accident tolerant fuel cladding material for next generation nuclear reactors. Grain refinement may lead to FeCrAl alloys with improved mechanical properties. However, the addition of Al retains the ferrite phase and makes the grain refinement difficult through conventional thermomechanical treatments. In this study, a model FeCrAl alloy, C35M, was processed by surface mechanical grinding treatment. Microscopy studies show the formation of gradient microstructures, consisting of surface nanolaminate layer and ultra-fine-grained sublayer. In situ micropillar compression tests reveal the gradient FeCrAl alloy has a high flow stress, exceeding 1.4 GPa, and excellent work hardening capability. A modified Kocks-Mecking model was utilized to explain the grain size-dependent work hardening behavior. This study sheds lights on the design of gradient ferritic steels with high-strength and good work hardening ability for various industrial applications.

Fast Neutron Irradiation Influence Analysis in Thermal-Hydraulics Aspect of HTR-10 Reactor Using Modified PEBBLE Program

Pebble Bed Reactor is a Gen-IV reactor that uses spherical fuel loaded into an active core. This reactor uses graphite that can be found on fuel and reflectors as a neutron moderator. The fast neutron can damage the atomic structure of fuel because the neutron with high energy can displace the graphite lattice. As its effect, the value of fuel thermal conductivity is influenced. PEBBLE is the program used for thermal-hydraulic calculation for pebble bed reactor type (PBR), and mPEBBLE was developed based on PEBBLE. The main difference between the two is that the mPEBBLE can take into account the fast neutron factor in the calculation. In this study, the thermal-hydraulics calculation for HTR-10 was performed using the mPEBBLE code. The result obtained is compared to the results from two different programs; THERMIX and WIMSTER. The comparison between the mPEBBLE and THERMIX has a temperature difference, which is relatively high at 63.3°C at the near-wall position. For the comparison between mPEBBLE and WISMTER, the noticeable difference happens in pebble center temperature, which is 65.3°C for the temperature difference.