Fast Neutron Reactors Research Articles

From “Inherently Safe” to “Proliferation Resistant”: New Perspectives on Reactor Designs

A recent American “mini-series” on Chernobyl, widely watched across the world, presented viewers with the concluding finding that this massive accident had occurred because the reactor design had inherent flaws; flaws that were known but not previously fixed because it was “cheaper” that way. The reactor design in question is the RBMK, and this paper will argue that this design was far from “cheap,” neither then nor now, and that its adoption as the second standard design for the Soviet Union’s nuclear power reactor fleet was based on much more than economic considerations. With the benefit of hindsight, it is easy to forget that reactor designs are always chosen for a multitude of reasons and never solely based on their technical or economic merits. Based on archival research, interviews, and industry publications, I show that approving and building RBMK reactors made good sense at the time, despite later claims to the contrary. Then I take the examples of a small modular reactor (SMR), the proposed NuScale Power Module, and a fast neutron reactor, TerraPower’s proposed Traveling Wave Reactor, to argue that we witness comparable negotiations today, as new designs for reactors (1) attempt to fit into existing safety and regulatory frameworks, (2) navigate security and nonproliferation concerns, and (3) embody visions of a specific sociotechnical order. I conclude that technical designs never occur in a socioeconomic, political, or cultural vacuum; instead, they are developed by people steeped in social norms, regulatory concerns, and economic expectations of a specific time and place. In the spirit of making this point relevant to practitioners, I will suggest ways of making these implicit frameworks visible, to actively and consciously start tweaking them, while staying aware of the implications that technical choices may have on our social expectations and vice versa.

Optimization of refueling times in fast neutron reactors

The importance of solving issues of increasing energy efficiency and energy saving for the development of the Russian economy was emphasized. The main directions of increasing the energy efficiency of the nuclear power industry (increasing the efficiency of using primary fuel and reducing the shutdown time of the power unit for repair and reloading of nuclear fuel) are highlighted. The importance of route optimization in solving applied problems of nuclear power is updated, in particular, in minimizing the time of nuclear fuel reloading, increasing the utilization factor of the installed capacity of NPP power units. The necessity of periodic reloading of nuclear fuel in reactors has been substantiated. The features of nuclear fuel reloading in thermal and fast neutron reactors are presented. The necessity of using in fast neutron reactors a system for guiding a capture on a fuel assembly using eccentrically located rotary plugs is substantiated. The necessity and importance of optimization of the process of guidance and movement of the overload mechanism in order to reduce the time to stop the power unit is shown. Mathematical models for the mechanisms of reloading fuel assemblies have been built. Time-optimal algorithms for the operation of overload mechanisms with two and three rotating plugs are proposed.

On the possibility of obtaining ultra-supercritical steam parameters at Nuclear Power Plants with fast neutron reactors using non-nuclear steam superheating

Nuclear and thermal power plants constitute the largest part of Russia’s power grid (about 67%). Today the world thermal power industry has already begun the transition to ultra-supercritical steam parameters, which makes it possible to increase the efficiency and reduce fuel consumption, and, accordingly, the discharge of harmful substances into the environment. Nuclear power plants (NPPs) need to increase their performance to keep up with other electric energy producers in the market. One of the ways to increase the energy efficiency of nuclear power plants with fast reactors is to improve the thermodynamic cycle. The paper presents computer-simulated steam-cycle arrangements for NPPs with a BN-1200 reactor using non-nuclear superheating of steam, as well as steam compression to obtain super-supercritical steam parameters, and an assessment of the efficiency of using these cycle arrangements.

Neutronography of Irradiated Reactor Austenitic Steels

The neutron diffracttion studies of the samples of fuel element claddings made of EK-164 and ChS-600 austenitic steels in the initial state and after operation in a BN-600 fast neutron reactor are reviewed. The samples were investigated in a wide range of fast neutron doses and irradiation temperatures, up to ∼80 displacements per atoms (dpa) and 628°C, respectively. Various microstructural characteristics, such as microstresses, dislocation density, and crystallographic texture are determined using full-profile analysis of neutron diffraction data. At high doses studied in this work, the irradiation temperature is the dominant factor responsible for the dislocation density. The procedures of analyzing and revealing defect types were preliminary developed on specially prepared nickel samples and austenitic alloys upon irradiation in the IVV-2M reactor core at a temperature near 80°С, which provided a low thermal mobility of lattice defects. The results obtained in some first diffraction studies on samples after their operation in the reactor core show that neutron data can be used to characterize the microstructures formed due to irradiation and, at least, estimate the mechanical properties of irradiated materials without their destruction and with no radiation risk for personnel.

Passive and Walk-Away Safe Small and Microreactors for Electricity Generation and Production of Process Heat for Industrial Uses

Abstract This paper presents the passive design and safety features as well as the biological dose estimates for end-of-life handling of postoperation for the walk-away safe, scalable liquid metal-cooled small modular (SLIMM) and very-small, long-life, modular (VSLLIM) small and microreactor with high-assay low enrichment uranium (HALEU). These passive operation sodium cooled, fast-neutron spectrum reactors have recently been developed at the University of New Mexico's Institute for Space and Nuclear Power studies. The SLIMM-small modular reactor could generate 100–10 MWth continuously, 24/7 without refueling for 6.3–68 full power years (FPYs), respectively, while the VSLLIM microreactor could generate 10–1.0 MWth without refueling for 5.8–92 FPYs, respectively. These reactors would be factory constructed, assembled, and sealed and then transported to designated sites by rail, a truck, or a barge. At the site, these reactors would be installed below ground, to protect against impact by an airplane or a missile, and mounted on seismic isolation bearings, to resist earthquakes. The SLIMM and VSLLIM reactors could provide for both electricity generation and process heat and the electrical plants can easily be connected to a central or distributed grid with renewable energy sources. The integrated VSLLIM reactor and other plant components for electricity generation and/or process heat production could all be deployed on a single railroad car, a barge, or an 18-wheeler truck, to small and remote communities, and advance bases or outposts.

Neutron diffraction study of samples of fuel element claddings made of austenitic steel

The paper presents the results of investigation of the performance of industrial fuel elements operating in the active zone of a fast–neutron reactor BN-600. As a continuation of previous works, we have studied fuel elements made of austenitic steel EK-164 after their irradiation under a maximal dose of 72.3 dpa at temperatures ranging from 370 to 628 °C. As in the earlier work on studying spent fuel elements fabricated from standard steel ChS-68, neutron-diffraction method was used that demonstrates a number of advantages over traditional methods, in particular, it is the least-exposing workers to radiation hazards and most suitable for studying bulk irradiated materials. Based on the experience gained, some quite typical microstructure distortions arisen under irradiation were revealed, changes in the density and type of dislocations depending on the dose and temperature were determined, and the dominant role of temperature at high doses of fast neutrons was confirmed. The behavior of EK-164 steel samples selected from different series is shown to depend on their initial structural state.

Simulating the Production of Medical Radioisotopes in a Fast Thorium-Cycle ADS with SERPENT

Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced continuously and transported as quickly as possible to the medical units where they are used. Neutron-rich medical radioisotopes are generally produced in research reactors, like technetium-99m, lutetium-177, holmium-166 and iodine-131. On the other hand, proton-rich radioisotopes are produced via reactions with charged particles from accelerators like fluorine-18, gallium-67, iodine-123 and thallium-201. Beside this, innovative nuclear reactors are advocated as solutions to the issues of nuclear waste production and proliferation threats. Fast neutron, thorium-cycle and accelerator-driven subcritical (ADS) reactors are some of the most promising of them, proposed as safer fuel breeders and “waste burners”. This article examines the use of a fast thorium-cycle ADS with liquid lead-bismuth eutectic coolant for the production of molybdenum-99/technetium-99m and lutetium-177. Burnup simulation has been made with the Monte-Carlo (MC) code SERPENT. It is demonstrated that MC codes can advantageously be used to determine the optimal irradiation time for a given radioisotope in a realistic reactor core. It is also shown that fast thorium-cycle ADS is an economical option for the production of medical radioisotopes.

MULTIPHYSICS MODELISATION OF AN UNPROTECTED LOSS OF FLOW TRANSIENT IN A SODIUM COOLED FAST REACTORS USING A NEUTRONIC-THERMAL-HYDRAULIC COUPLING SCHEME

Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.

Physico-chemical Modeling of Phase Separation in Fe-21.4 Cr Steel with 1.14 Mo Using the Provisions of Non-equilibrium Thermodynamics

The use of ferritic-martensitic steels, for example, Fe-Cr and Fe-Cr-Al alloys as structural material for fast neutron reactors has been advocated due to their relatively low rate of swelling at elevated temperatures. Because of the high Cr-content, the Fe-Cr alloys are not suitable for use at temperatures around 500°С due to the miscibility gap in the Fe-Cr system. Phase separation in such alloys can also be prevented by doping with elements that impede the segregation of chromium. In solid state physics for calculating phase separation in similar metal systems are used physicochemical and phase-field modeling. The aim of the work is the physicochemical modeling of diffusion phase transformation and determination of the long-term microstructural stability of the Fe-21.4 Cr-1.16 Mo. A conventional Fe-21.4 Cr alloy is used as a reference material. The article proposes an integral approach to modeling phase separation in chromium alloys, combining the determination of diffusion coefficients and fluxes of elements, taking into account their dependences on the concentration and an assessment of the mutual diffusion of elements, using the provisions of nonequilibrium thermodynamics. The calculated values of diffusion fluxes are used to calculate the current concentrations of carbon and chromium in the alloy and the size of chromium formations. They show that the thermal stability of the Fe - 21.4% Cr alloy with 1.16% Mo is much higher than without molybdenum. In alloy Fe – 21,4% Cr – 1,16% Mo at a temperature of 973°K, the chromium concentration during the same operation time decreases three times slower with the formation of inclusions of the σ-phase about 6 microns in size.

Взаимодействие изотопов водорода с ферритно-мартенситной сталью ЭК-181 (Русфер). Обзор полученных результатов

Ferritic-martensitic steel with rapid decay of activity EK-181 (Rusfer-EK-181, Fe – 12 Cr – 2 W – V – Ta – B) was developed in Russia (A.A. Bochvar Institute) as a structural material for use in the active zones of fast neutron reactors, as well as in thermonuclear and hybrid reactors. The team of authors has been studying various aspects of hydrogen interaction with Rusfer steel for a number of years, primarily retention, diffusion, and the effect of various damages on capture. The paper summarizes the results of the research performed by the authors.

Достижимость радиологической эквивалентности в ЗЯТЦ на базе БР с учётом факторов неопределённости сценариев развития ядерной энергетики в России до 2100 г. Часть 2. Миграция радионуклидов

The potential health risk of future generations of people from radioactive waste (RW) disposed in deep repository is estimated for three scenarios for the development of nuclear power in Russia. There is currently a zone of uncertainty when making decisions about future power generation technologies. For this zone of uncertainty, the following three scenarios are considered: 1) appli-cation of fast neutron reactors, FR, only; 2) application of thermal neutron reactors, TR; and 3) ap-plication of FR and TR in combination. Long-lived RW from FR is assessed to be ten times less toxic than RW from TR. After reprocessing of the wastes to extract 90% of uranium and plutonium for incineration and americium for transmutation in FR, the toxicity of RW is again reduced by 10 times. Committed effective doses and lifetime attributable risk (LAR) to the public are estimated with account of radionuclides migration from the deep storage to the surface. Due to the RW re-processing the time to the achievement of radiation equivalence of RW stored in the repository and natural uranium ore is reduced to an acceptable time, at which the integrity of RW packages is guaranteed. From the standpoint of modern standards of radiological protection and minimiza-tion the potential carcinogenic effects of radiation exposure on the population, priority in the de-velopment of nuclear energy in Russia should be given to the first scenario of the development of nuclear power, in which FR are used as much as possible to generate electricity. This conclusion should be taken into consideration in adjusting the Strategic plan for the development of nuclear power in Russia.

A hybrid approach for neutronics calculations in the neutron shielding of sodium fast reactors

Neutron calculations in the neutron shielding of fast neutron reactors are a complex problem as deterministic schemes are usually not suited for such calculations while Monte-Carlo codes have poor computational performance due to the very low flux levels in neutron shields. In this article, both methods are studied, as well as a hybrid scheme on the neutron shielding of the ASTRID fast reactor benchmark. This hybrid scheme uses a fission source calculated by a deterministic code in order to precisely calculate neutron fluxes in the shielding with a Monte-Carlo code using variance reduction techniques. This provides reference results in order to validate deterministic calculations. Comparisons between deterministic codes and this hybrid reference show that large biases are obtained, up to 50%. Further studies are made to reduce the biases, showing that many physical phenomena should be treated, including anisotropy of the scattering law at high energies and spatial self-shielding inside the boron carbide shielding. These improvements reduce the biases to less than 10%. Finally, some applications to designing criteria for the neutron shielding are presented, such as gas production in the neutron shielding and activation of secondary sodium at the intermediate heat exchanger (IHX).

ТЕПЛОГИДРАВЛИЧЕСКИЕ ИССЛЕДОВАНИЯ ВЫСОКОТЕМПЕРАТУРНОЙ ЯЭУ ДЛЯ ПРОИЗВОДСТВА ВОДОРОДА И ДРУГИХ ТЕХНОЛОГИЧЕСКИХ ПРИЛОЖЕНИЙ

An important problem determining the development of clean energy is the involvement of hydrogen in the fuel cycle. At present, the main method of hydrogen production is steam methane conversion. In the long term, large-scale hydrogen production, this method is not viable due to the consumption of non-renewable resources and the emission of greenhouse gases. Alternative methods of hydrogen production by water splitting methods using thermochemical or electrolysis processes require a high-temperature heat source. Nuclear reactors can serve as the most widely used high-temperature heat sources. The performed neutron-physical and thermophysical studies have shown that there is a fundamental possibility to provide the required parameters of a high-temperature (900-950 °C) with a 600 MW (thermal) fast neutron reactor with a sodium coolant for hydrogen production. It’s possible on the basis of one of the thermochemical cycles or high-temperature electrolysis with a high coefficient of thermal utilization of energy. It is shown that the temperature regime of core fuel elements is determined by a large number of parameters that have a regular and statistical nature. The developed methodology and numerical program allows to take into account, in the fuel assemblies shaped during the campaign, the effect on the temperature distribution of the fuel element cladding and temperature irregularities along the fuel element perimeter in the interchannel exchange fuel assembly, the random distribution of channel cross-sections and the heat generation of fuel elements using the Monte Carlo method, also other factors. For various reactor operating regimes, zones with stable temperature stratification with large gradients and temperature fluctuations have been identified. The results obtained make it possible to judge the amplitude and frequency characteristics of temperature pulsations in these potentially dangerous areas. The relative small size, the type of coolant, the choice of fissile material and structural materials make it possible to create a reactor with inherent properties that ensure increased nuclear and radiation safety.

Laser cladding of nickel base hardfacing material: Material analyses and manufacturing process evaluation on a scale one demonstrator

In fast neutron reactors, contact areas of moving parts usually require cobalt-free hardfacing coatings, as cobalt is highly activated under neutron flux. This is particularly critical for the insert holes of the diagrid for the positioning of the hexagonal fuel tubes that have to be internally coated. In this article, we propose to present the development of the cobalt-free hardfacing material up to the manufacturing of the inner clads with a specific deep laser cladding nozzle. In previous presentations, laser cladding has been identified as a deposition process that could increase the performances of the hardfacing materials compared to the standard process (Plasma Transferred Arc Welding). In parallel, the potential interest of some nickel base materials such as Colmonoy® 52 or Tribaloy® T700 has been demonstrated. Unfortunately, the deposition of these fragile alloys requires a preheating of the substrate over 450 °C. More recently, Nucalloy® 453, a new hardfacing nickel base alloy has been evaluated and demonstrated simpler deposition conditions that requires lower preheating temperature (<300 °C). The article presents the evaluation of Nucalloy® 453 with material analysis and wear tests. The microstructural characterization is compared to Colmonoy® 52, which is a similar NiFeCrSiBC alloy. Finally, the laser cladding of a scale one demonstrator is presented: two inner zones of a 1 m cylinder of 100 mm diameter are laser cladded, thanks to a deep cladding nozzle. The advantage of the laser cladding process is compared to Plasma Arc Transferred Welding.

Processing of fast neutron reactor fuel by electrorefining: Thematic overview

This work provides basic knowledge on the spent fuel management on the basis of the published literature data on electrorefining. This review examines three main areas of work devoted to electrorefining. These are electrodeposition and electrodissolution using solid and liquid electrodes, as well as mass transfer in phases present during electrorefining. As part of this research, the composition of the irradiated metallic fuel was estimated. Due to the great potential difference between solid cathodes, it is possible to separate actinides from lanthanides. The co-deposition of metallic Pu and U in the eutectic LiCl-KCl melt containing UCl3 and PuCl3 indicates stable co-precipitation of U and Pu at U3+ concentration less than ~0.2 wt. Periodically performed electrical transport of ions to liquid (Cd) and solid cathodes in the galvanic mode made it possible to deposit preferentially U on the solid cathode, and Pu and Am on the liquid cathode. Oxidation of these metals caused fluctuations in the anode potential. The electrode processing after electrorefining is investigated. This process consists of oxidizing the actinides remaining in the liquid electrode by adding CdCl2 and removing the associated chloride by high-temperature distillation. During the electrorefining of irradiated metallic fuel, the fission products accumulate in the molten salt. Reduction of uranium on a solid cathode from a spent molten salt using a liquid Cd-Li anode is considered. A model that describes electrorefining with a liquid metal anode, solid cathode, and molten LiCl-KCl salts, is presented. The formation of plutonium at the surface of a solid cathode is analyzed. In a one-dimensional model of an electrorefiner, it is shown that the concentration of Pu at the cathode cannot be predicted from the Cm concentration in the melt.

The Use of Thorium in a Nuclear Power System with Thermal and Fast Neutron Reactors

The concept of a fast neutron reactor with a sodium coolant with metallic U–Pu fuel in the core and metallic uranium and thorium in the blankets is considered. The achievement of the required system characteristics for the initial loading and surplus fuel breeding is demonstrated. Several scenarios of the development of a two-component nuclear power system in Russia based on the VVER-type thermal reactors and fast reactors with metallic fuel involving the thorium resources are considered.

Protective Coating of Irradiation Device for Fast-Neutron Reactor

The objective of this paper is to select a coating on the surface of zirconium hydride (ZH) in order to reduce hydrogen losses under the operating conditions of an irradiation device of FN-reactors. Tests were carried out on the samples of the moderator units made of zirconium hydride. Oxide, phosphide-oxide and oxide-enamel coatings were considered as a protective layer. Hydrogen yield out of the ZH was studied using a calorimeter and a mass spectrometer by thermal heating of the samples according to the test program (heating rate of 10 °C/min, maximum temperature of 1000 °C, isothermal exposure of 40 min, cooling with a furnace). The hydrogen content in the samples was determined by high temperature vacuum extraction. It is shown the oxide coating is the most technologically advanced and effective. A thermogravimetric analysis of the ZH sample with the protective oxide coating was carried out, and the temperature-time dependence of a stoichiometric coefficient (H / Zr) was determined. The amount of the hydrogen released was determined in a temperature range of 20–1100 °C. The temperature increase in up to 1100 °C leads to almost total loss of hydrogen (98–99 %). In this case, the sample retained its integrity, and the coating exfoliated and fragmented.

Research on structure selection and design of LBE-cooled fast reactor main coolant pump

Fast neutron reactor has become the preferred reactor type for advanced nuclear energy systems in the world, representing the development direction of the Generation IV nuclear energy system. The lead-cooled fast reactor (LFR) uses pure Pb/LBE as coolant, the natural characteristics of lead alloy are combined with the structural design of LFR, which constitutes the inherent safety characteristics of LFR and the system. In this paper, the main coolant pump in the Generation IV of LFR is taken as the research object. According to the design requirements of the main coolant pump in LFR fast reactor, the hydraulic characteristics of the main coolant pump with double-outlet volute type and space guide vane type under normal operation and accident operation are compared and analyzed. It is found that, under normal pump conditions the hydraulic performance of double-outlet volute type main coolant pump is a slightly better, but the space guide vane type has obvious advantages in the ability to deal with accident conditions and the better reliability in its long-term operation. Finally, the space guide vane structure is determined as the design program of the main coolant pump of CLEAR-I fast reactor.

Conception of delayed neutron detector blocks in a sodium cooled fast reactor

The primary vessel of a sodium-cooled fast neutron reactor (SFR) must remain free of any contaminant released by a cladding failure in a fuel assembly. A comprehensive detection system of cladding failures has then to be considered right from the beginning of the reactor design. One sub-system, called delayed neutron detection (DND), aims to search for delayed neutron precursors (DNP) through an ex-vessel analysis of a sodium sample, which goes to a container, usually named a pot, surrounded by neutron detectors. This paper presents a methodology to optimize the hydraulic part of the design of the DND detection block. From a detection scenario that identifies the physical parameters in play, a computational model is derived, which couples fluid mechanics and neutron transport. Different types of pot designs are considered, for which the sodium flow velocity, DNP concentration and detector activities are assessed. A new type of pot with a coil shape is proposed and shown to have the better performances with respect to the detector response to delayed neutrons.

A Study of the Radiation Tolerance of CVD Diamond to 70 MeV Protons, Fast Neutrons and 200 MeV Pions.

We measured the radiation tolerance of commercially available diamonds grown by the Chemical Vapor Deposition process by measuring the charge created by a 120 GeV hadron beam in a 50 pitch strip detector fabricated on each diamond sample before and after irradiation. We irradiated one group of samples with 70 MeV protons, a second group of samples with fast reactor neutrons (defined as energy greater than MeV), and a third group of samples with 200 MeV pions, in steps, to () × 1015 protons/cm2, () × 1016 neutrons/cm2, and () × 1014 pions/cm2, respectively. By observing the charge induced due to the separation of electron–hole pairs created by the passage of the hadron beam through each sample, on an event-by-event basis, as a function of irradiation fluence, we conclude all datasets can be described by a first-order damage equation and independently calculate the damage constant for 70 MeV protons, fast reactor neutrons, and 200 MeV pions. We find the damage constant for diamond irradiated with 70 MeV protons to be 10−18 cm2, the damage constant for diamond irradiated with fast reactor neutrons to be 10−18 cm2, and the damage constant for diamond irradiated with 200 MeV pions to be 10−18 cm2. The damage constants from this measurement were analyzed together with our previously published 24 GeV proton irradiation and 800 MeV proton irradiation damage constant data to derive the first comprehensive set of relative damage constants for Chemical Vapor Deposition diamond. We find 70 MeV protons are 2.60 ± 0.29 times more damaging than 24 GeV protons, fast reactor neutrons are 4.3 ± 0.4 times more damaging than 24 GeV protons, and 200 MeV pions are 3.2 ± 0.8 more damaging than 24 GeV protons. We also observe the measured data can be described by a universal damage curve for all proton, neutron, and pion irradiations we performed of Chemical Vapor Deposition diamond. Finally, we confirm the spatial uniformity of the collected charge increases with fluence for polycrystalline Chemical Vapor Deposition diamond, and this effect can also be described by a universal curve.