The Advanced Lead Fast Reactor European Demonstrator (ALFRED) is a pool-type lead-cooled fast reactor demonstrator of up to 300MW(th). The purpose of ALFRED is to demonstrate the viability of reliable electricity production of a European Lead Fast Reactor (LFR) system as a next-generation Small Modular Reactor (SMR) based commercial power plant. ALFRED is cooled with pure lead in forced...
In fast breeder reactors, top shield acts as reactor cover in the axial direction and consists of Roof Slab and rotatable plugs. The rotatable plugs, which are mounted over large diameter bearings are rotated to facilitate handling of fuel subassemblies. Since the interface between the rotating and stationary members of the top shield forms boundary for primary radioactive Argon cover gas, the...
In a previous study, it was reported that it is possible to establish Breed and Burn fast reactors with rotational and spiral fuel shuffling scheme using metallic fuel and lead-bismuth coolant. In the rotational shuffling scheme, fresh fuel assemblies composed of natural uranium are loaded at the edge of core and approach to the center by each shuffling and move to outer again. In the spiral...
Abstract
There are a huge data volume based upon theoretical and experimental metallic fuel studies in the fast spectrum nuclear installations. A higherdensityof such fuel allowed to get a high breeding value has a considerableresearch interest. But there are number ofrestrictions not allowed the practical implementation of the high power reactor projects with such fuel at that moment. ...
The industrial energy complex with fast neutron reactor is on the pre-design stage. A conceptual design (‘shape design’) of the industrial energy complex has been issued in 2019. The main solutions and technical and economic indicators, the perspective ways of further developments and criteria were defined. In 2020, the Industrial energy complex' works have moved to the stage of a feasibility...
A dual-purpose fast reactor with a light / heavy liquid metal coolant and a large-scale production of Pu-238 is considered. A universal target complex for large-scale production of Pu-238 is located inside the reactor. Np-237 is considered as a starting material for the production of Pu-238. The target complex has a heterogeneous structure, including Np-237 and a moderator with a high atomic...
Nuclear energy utilization for hydrogen production is experiencing a growing momentum worldwide, associated with the unprecedented interest in building large-scale hydrogen production plants to support national and international decarbonization and climate change mitigation plans. Thermochemical water splitting cycles coupled to nuclear power plants are one of the sustainable solutions to...
Among the various types of liquid metal cooled fast reactor, lead cooled fast reactor (LFR) can take a closed fuel cycle to manage fertile fuel and actinide efficiently. LFR can adapt an External Boiling Bayonet Steam Generator (EBBSG) system instead of conventional once-through high pressure steam generator. LFR with the EBBSG has advantages in terms of thermal efficiency and safety during...
In recent years, interest in (very) small and medium size reactor– (v)SMR – concepts for specific purposes has grown. They are characterized by irregular geometries. Though Monte Carlo methods for safety assessment are becoming more and more standard for steady state simulations, they are not yet mature enough for transient applications. To perform future coupled transient safety assessments...
The neutron lifetime is an important parameter of the reactor kinetics. When the inserted reactivity is more than the effective fraction of delayed neutrons, the reactor kinetics becomes very rapid. The fast reactor kinetics can be slowed down by increasing the neutron lifetime. The possibility of using lead isotope 208Pb as a neutron reflector with specific properties in the lead-cooled fast...
This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid lithium cooled fast reactor core, for a Thermal Propulsion Engine, combined with a simple neutron population computing for an infinite cell lattice. Two types of fuel configuration were studied: pin cell assembly and matrix fuel assembly, with all requirements regarding safety conditions...
The European-wide research on the Lead-cooled Fast Reactor (LFR) has been steadily advancing the technology readiness level at a point where, for further targeting industrial maturity, the need of a demonstrator has risen. The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED), along with its support research infrastructure, is the missing link of the innovation chain being...
We have investigated main neutronic parameters and core performance parameters of the reference China Initiated Accelerator Driven System (CiADS) core after loading uranium nitride instead of uranium oxide. Based on neutronic calculations performed with the Serpent code, only 15.2% enrichment is needed to provide similar sub-criticality as the reference design when using UN instead of oxide...
If the procedure can be decomposed and the complexity of the evaluation procedure can be quantified, it is an effective way to reduce the operator's workload and reduce the human error rate. In this study, a measure is developed, based on the task complexity (TACOM) evaluation method, combined with the characteristics of the China Experimental Fast Reactor. The TACOM measure consists of five...
The operation experience of sodium-water steam generators has demonstrated that the main source of hydrogen ingress into the secondary sodium is the process of hydrogen depolarization (equation 2) of the cathodic electrochemical corrosion process (equation 1) on the steam-generating steel surfaces from the water side:
3Fe + 4H2O = Fe3O4 + 4H2 (1)
2H2O+2e- = 2Hkt + 2OH- (2)
Hydrogen...
Aiming at the 600MW demonstration fast reactor（CFR600）activation method experiment, this article introduces the comprehensive experimental measurement system and the basic flow of the experiment, describes the experimental principle and the expected experimental content, and focuses on the design of the CFR600 activation method experiment. First, according to the project schedule, the...
Remote Handling is one of the prominent areas of research and development for the nuclear sector for the remote manipulation of irradiated nuclear fuels and structural materials. The standard coding for the Remote handling devices for radioactive materials is covered as part 1: General requirements under ISO: 17874-1:2004 & part 3: Electrical master-slave manipulators (Servo Manipulator) under...
Currently, the world society recognizes serious environmental problems with anthropogenic impact on the environment and almost does not dispute the main source of its pollution. However, the hope that growing demand of world economies for electricity can be provided by so-called alternative energy sources - solar and wind installations is illusory.
At the same time, the world's existing...
The development of uranium-plutonium compounds as nuclear fuel is a new nuclear technology using poorly studied highly active uranium compounds. The toxicity and consequences of the biological action of such compounds are the subject of this work.
It should be noted that during the production of MNUP fuel there is a possibility of impact on the population and the environment, which is mainly...
Public concerns about nuclear safety and environmental impact of the operations of nuclear power plants and accident scenarios, have spurred innovative research into the study of advanced reactor fuels with the fast breeder reactors belonging to the new age generation reactors.
These reactors termed Gen IV reactors are evangelized for their many positives specially to mitigate the issues of...
Currently the largescale nuclear power system, based on the two-component technology platform has been developed in Russia. The technology implies the buried spent fuel to be safe when its and natural uranium impacts on human health to be equal.
To 2100 fast reactors (FR) and thermal reactors (TR) will consume 541.7 thousand tons of natural uranium, due to the reactors operation 7.523...
Measurement of gas flow rates in exhaust ducts for radioactive plants are essential from process as well as radiation safety requirements. Averaging Pitot Tube (APT) is mechanical device which generates differential pressure proportional to flow rate when inserted in the path of the flow. In this paper we discuss the reason for the selection of this particular type of flow measurement...
Hydrogen and hydrogen technology are expected to have a key role as an energy carrier for the technic and economic systems. This are expected to be a new impulse for the nuclear’s integration in the grid.
Large-scale demonstration projects of the low carbon hydrogen production require major investments by countries and long term strategies. Taking into account the specific of nuclear...
Tilting of reactor vessel and other components due to circumferential temperature gradient is a critical safety and operational issue in loop type Fast Reactors. Natural convection of reactor cover gas developed in the narrow component penetrations is the main cause for such a phenomenon. Numerical studies have been carried out using CFD code to investigate the possibility for development of...
The report will present the results of assessing the risk of the impact of a complex of negative factors of production of MNUP-fuel of a radiation and non-radiation nature on the health of workers. Such factors include both the combined effect of external photon and neutron irradiation due to the inhalation of 238U, 239Pu and their decay products, and the toxic effect of the compounds included...
This article provides the results of a case study on a comparative evaluation and ranking of six power generation technologies including 2 nuclear, 2 fossil fuel and 2 renewable power generation options. The performed analysis was based on the combination of the approaches proposed within the project New Energy Externalities Development for Sustainability (the NEEDS project) and the INPRO/IAEA...
Transition to the system of the closed nuclear fuel cycle based on reactors on thermal and fast neutrons is an important stage of formation of two-component energy industry, which provides enhancement of efficiency of uranium resource efficiency. The closed nuclear fuel cycle at the international level can be presented both from the viewpoint of energy security and stability and from the...
Development of ultrasonoscopy system “Vizus” is aimed at increase of operation safety of power units with sodium cooled BN-type reactors. “Vizus” system application permits to detect foreign objects in the space above the core and, as a result, to prevent failure of devices passing through the CPS column and core elements at rotary plug rotation.
The system is intended for registration and...
The heavy liquid metal coolant technology (HLMC) is a set of measures that make it possible to do the following:
- to prepare the coolant for filling the primary circuit of the reactor facility (RF);
- to maintain the conditions in the coolant so as to ensure corrosion resistance of structural steels;
- to perform purification of the coolant from solid-phase slag based both on lead oxides...
Development results of submerged electromagnetic pump (EMP) on liquid lead for reactor BREST-OD-300 are presented. EMP is planned to use for liquid lead level regulation in the reactor during putting it into exploitation under partial or full EMP submerging in lead. Main EMP parameters: pressure head 1.0 MPa, nominal flow rate 2.0 m3/hr, lead temperature 390-420 °С. Required service life...
Sodium, owing to its high heat transfer properties and excellent compatibility with structural materials is the preferred coolant for Liquid Metal cooled Fast Breeder Reactors (LMFBR). Apart from its favorable properties, sodium also poses a concern due to high chemical reactivity with air. Accidental sodium leaks from secondary circuit may result in fire due to sodium reaction with oxygen and...
Results of heat transfer analysis and service life time estimation are presented for submerged electromagnetic pump (EMP) on liquid lead for reactor BREST-OD-300. EMP is planned to use for liquid lead level regulation in the reactor during putting it into exploitation under partial or full EMP submerging in lead. Main EMP parameters: pressure head 1.0 MPa, nominal flow rate 2.0 m3/hr, lead...
Breed and Burn fast reactors are very attractive reactor concept. It can achieve very high burnup without reprocessing facility using depleted uranium or natural uranium as the fuel for the reactor. To achieve criticality in the equilibrium burnup condition improvement of neutron economy is the essential issue. The feasibility of the reactor can be discussed based on the neutron balance of a...
The breed and burn fast reactor concept have been a familiar with nuclear engineers since the 1950s in which the depleted or not enriched fuel is loaded in a core and breed the fissile fuels while burning them. Development of Breed-and-Burn (B&B) reactor means of effective utilization of uranium resource and reduction of spent fuel. There are issues to be solved as of irradiation damage in...
A 12% Cr ferritic/martensitic (F/M) steel, HT-9, has been used as a primary core material for nuclear reactors. Welding is inevitably used in the nuclear structure design. Fusion welding processes such as gas tungsten arc welding (GTAW) were broadly applied for the F/M steel components. Unfortunately, in the fusion zone of F/M steel weldment, delta (δ)-ferrite is frequently produced, which is...
The harsh neutron irradiation environments in the core region of fast breeder reactors (FBRs) pose a unique challenge for cladding materials. Microchemistry and Microstructural changes resulting from displacement damage and creep rupture are anticipated for structural materials after extended neutron irradiation. Various irradiation effects on the service performance of cladding materials need...
During their service life, fast neutron reactor oxide fuel elements operate at high neutron fluence rate, high linear power, and high temperature environment. The fuel elements will exhibit complex irradiation -thermo-mechanical coupling characteristics. The stress-strain analysis calculation is very important for the design of fast reactor fuel elements. This article focuses on the...
Three methods are recommended for determining the corrosion coefficients of steels in liquid metal (LM).
The essence of the first method, the so-called metallographic method is to measure the thickness of oxides and study their structure and composition. The main advantage of the method is its relative simplicity, as there is no need to remove any residues of LM or oxides from the specimen,...
Possibility of measurement of oxygen concentration in sodium discussed in the papier. Usually by means of the plug indicator “plugging temperature” is measured and supposed it is equal saturation temperature of sodium with impurities. Then, concentration of oxygen is determined using equation for solubility of oxygen in sodium. As a rule, Noden´s equation is used. But solubility of oxygen in...
The state of high enriched В4С (industrial and re-fabricated quality) consisting of emergency protection elements, irradiated up to neutron fluence 7x1022 cm-2 (Е>0.1 МeV), was investigated. The maximum burnup of an 10В isotope at two-year irradiation period has made 25 %. That burnup decreases on diameter and height of an absorber column. High local mechanical deformation of emergency...
CEFR is a fast neutron experimental reactor with the coolant of liquid metal sodium. The primary loop’s corridor of sodium purification system, as we know the name of room 309/1, is one of the sodium aerosol containment chambers, where laid out with many sodium pipes of high temperature, especially the pipe with high-level radioactive sodium inside which is come from reactor main vessel. In...
Based on the lessons learned from the TEPCO’s Fukushima Daiichi NPPs’ accidents, evaluation of fission products (FPs) transfer during the severe accidents is quite important in the regulation activities.
In core expansion phase caused by core disruptive accident (CDA) of sodium-cooled fast reactors (SFRs), a large bubble consisting of mixture of core materials and sodium vapor inside,...
The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (USDOE), Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled...
In the fast reactor, flow fields which have different temperature flow out of heads of assemblies and then mix intensely, which leads to temperature fluctuation. The temperature fluctuation in the fast reactor is very serious due to the thermal characteristic, especially in the upper plenum, which may cause thermal fatigue in the structure of fast reactor. In the published work, simplified...
A vital part of the licensing process for advanced (non-LWR) nuclear reactor developers in the United States is the assessment of the reactor’s source term. The source term represents the potential release of radionuclides from the reactor system to the environment during normal operations or accident sequences. While historically source term assessment for LWRs has followed a bounding...
Ensuring hydrogen safety is one of the most important conditions of the general substantiation of NPP safety.
The greatest danger of the release of a huge amount of hydrogen arises in relation to nuclear stations in water and water stations during the beyond design basis steam-zirconium reaction and in case of leaks between the second (sodium) and third (water) circuits of fast reactors with...
Mechanical consequence which might be caused by core disruptive accidents (CDAs) is one of the major concerns in fast reactor safety. Once core disruption occurs caused by severe re-criticality, core materials are dispersed azimuthally and radially. The dispersed materials, e.g., liquid/particle mixture of fuel and steel, penetrate into the pin bundles and control rod guide tubes (CRGTs), and...
The in-vessel source term problem is to determine the partitioning of radionuclide (RN) in fuel, coolant and cover gas following a hypothetical severe accident. The determination of the in-vessel source term is a challenging task, as there is significant uncertainty involved in RN transport. The RN transport mostly controlled by their diffusion characteristics in the fuel, chemical...
The Versatile Test Reactor (VTR) is currently under development by the U.S. Department of Energy (DOE). It will provide very high fast neutron flux irradiation capabilities that are currently unavailable in the U.S. Given the increasingly large number of advanced reactor concepts being pursued in recent years, this irradiation testing capability will be essential to support maturation of these...
A major appeal of sodium fast reactors is their passive safety capabilities. To demonstrate this, the Fast Flux Test Facility (FFTF) conducted a series of Loss of Flow WithOut Scram (LOFWOS) tests at up to 50% power. Experimental results from this test were made available through an IAEA CRP for use in a code benchmarking activity. In this work, the System Analysis Module (SAM) was used to...
According to the sodium cooled fast reactor operation experience, about 100 sodium leakage accidents have happened in history. Sodium fire is a typical hazard in SFR, which is also one of the main reasons for its unavailability, and may be one of the main contributors to the total reactor risks. Study on sodium fire PSA methodology can not only quantitatively evaluate the sodium fire risk in...
System dynamic simulation of loss of flow without SCRAM test carried out in FFTF has been carried out using plant dynamics code DYANA-P. DYANA-P has one-dimensional models for various sub-systems of sodium cooled fast reactor. Thermal models are based on heat balance between various sections exchanging heat. Hydraulic model is based on momentum balance between various flow segments in sodium...
Passive safety systems are used in generation III+ evolutionary reactors and in generation IV advanced reactor designs, especially for the decay heat removal following an accidental event. These systems allow with one or more loops the heat transfer from the primary system to the external environment through the natural circulation of fluids or through boiling and condensation phenomena. A...
The technologies of information retrieval in a database with full-text semantic indexing are considered. The information retrieval process is considered as a cognitive-oriented process. The semantic image of the document context is presented as an ontology. An ontology is defined as a set of three interconnected systems (functional, conceptual and terminological), on which the operation of...
Тhis work is carried out in order to assess the compliance of the radiation protection of personnel working at the complex experimental installations of JSC «SСhE» with the requirements of the national radiation safety standards to limit the generalized risk of potential exposure and the IAEA recommendations for not exceeding the control level of the minimum significant radiation risk.
On the...
Fast Breeder Test Reactor (FBTR) in India is designed for 40 MWt Thermal 13.2 MWe. At present FBTR is operating at 32 MWt with 56 fuel sub assemblies (FSA) of 48 Mark I and 8 MOX type fuel sub assemblies. Mark I FSA are of Pu-U Carbide fuel with 70% Pu and MOX FSA are of PuO2 (44%) and UO2 (56%). Due to constraint on minimum shut down margin of 4200 pcm, the core could not be expanded and...
High level liquid Waste (HLLW) storage tanks with large storage capacity weighing a few tens of MT are proposed to be used in Fast Reactor Fuel Reprocessing Plant (FRP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam to store the HLLW. Six years storage capacity is envisaged for allowing Ru106 to decay sufficiently before sending the HLLW for vitrification.These tanks have lot of...
The Eddy Current Flow Meter (ECFM) is a robust and reliable inductive sensor for measuring the flow rate of liquid metals. Since there is no direct contact between sensor and liquid metal, it can be used in chemically aggressive environments and at very high temperatures of up to 600 °C. This allows the ECFM to be deployed, for example, as part of the safety instrumentation in liquid...
Many elastomers seals are used in the nuclear industry. Among these elastomers, ethylene propylene diene monomer (EPDM) and silicone rubbers have excellent radiation stability. Both the rubbers can be used for gasket and O-ring application in Reprocessing Plants. To study the suitability of these rubbers for application in the plant, EPDM rubber compound and silicone rubber compounds were...
Here described a method for determination of mean prompt neutron lifetime in fast metal cores during critical experiments held in RFNC – VNIIТF using assembly machine FKBN-2. The evaluation of derivative using experimental dependence between asymptotic decrease coefficient and core parts gap α(Н) was proposed further to determination of the delayed critical state of the core. The value ...
The equipment supports are in constricted arrangement in the main vessel of the fast reactor. Under the condition of earthquake, equipment supports may sustain damage caused by the interaction between equipment supports and fluid, therefore, the evaluation of the fluid-structure interaction effect is an important aspect of the structural safety assessment of fast reactors. Using the added mass...
Characteristics of the molten-salt reactor-burner (MSR-burner) of minor actinides (MA), which are concentrated in spent nuclear fuel of power reactors, depend significantly on the physical-chemical properties of the fuel composition. In particular, the MA transmutation efficiency is mainly determined by the concentration of actinide fluorides in the molten-salt fuel composition [1]. In this...
In order to reduce the long-term potential hazard of waste from the reprocessing of spent nuclear fuel from thermal reactors and to increase the enviroment attractiveness of nuclear power in our country, work is underway to create a molten-salt reactor-burner of minor actinides (MSR-B). The first stage on this path is the creation of an investigative molten-salt reactor (IMSR) for testing key...
The report presents results of studies of the physicochemical characteristics of radioactive workplace aerosols formed during the production of mixed nitride uranium-plutonium fuel (activity particle-size distribution, nuclide composition, lung absorption type, elemental composition, reactive properties in the air). Taking into account these characteristics, the dose coefficients were...
The objective of this work is to reveal the dependences of the separation of americium-241 and uranium using sorption technology based on the solid-phase extractant TODGA. In technologies for the purification of radioactive waste of low and medium activity levels with low contents of actinides, sorption and ion exchange methods are widely used due to their high selective. The required...
The Versatile Test Reactor (VTR) is a reactor under development in the United States of America to provide a very high-flux fast neutron source. This reactor will accelerate the testing of advanced nuclear fuels, materials, and other potentially irradiated components. As this reactor design effort is underway to support eventual construction and operation, a necessary step is the development...
Characterization of alpha emitting nuclide and other fission products in the radioactive waste generated in reprocessing plants is a regulatory requirement for their disposal. The assay of plutonium in the solid radioactive wastes could be carried out either using gamma spectrometry or neutron counting, depending mainly on the surface dose of the container. Presence of large amount of fission...
In Russia, research is actively underway to develop a specialized molten-salt burner reactor (MSR-burner) of minor actinides (MA) from spent nuclear fuel of power reactors. Two candidate fluoride salts, LiF-BeF2 [1] and LiF-NaF-KF, are considered as the solvent of the reactor fuel components.
The purpose of the present paper is to study MA transmutation in the MSR-burner based on selected...
This paper aims at finding solutions of so important problems of nuclear power as decreasing the scope and the number of technological operations, as well as enhancing the proliferation resistance of fissile materials in nuclear fuel cycle by means of minimal changes in the cycle. The method is including fusion neutron sources with thorium blanket into future nuclear power system. In addition...
To determine uranium in the metallic phase in the presence of uranium oxide there is a reliable, so-called “bromine method”, which implies a metallic-oxide mixture treating in the bromine ethyl acetate solution. However, analogous manipulations with rare earth metals and their oxides do not provide such reliable data. Reduction melting of oxides in a graphite crucible with the melt composed of...
Programmable Logic Controllers offer complete automation solution and flexibility to control in a plant like the nuclear fuel reprocessing plant. To ensure the safety of both the plant and personnel, continuous monitoring and diagnostics of plant parameters are implemented through various means. Audible and visual alarms are provided to alert the operator in case of process abnormality. But...
Technology to utilize a dry recycled MOX powder has been developed as a part of MOX fabrication technology development for fast reactors. The purpose of this study is to develop a technology to control the density of MOX pellets with use of dry recycled MOX powder. A roll crusher and a jet mill were employed to prepare the recycled MOX powder which had three kinds of particle sizes (coarse,...
Melts based on the LiCl-KCl eutectic are becoming attractive in various industrial fields, including nuclear industries. However, their transport characteristics have not yet been sufficiently studied.
The purpose of this work is to study the electrical conductivity of melts similar to those formed during the dissolution of real nitride spent nuclear fuel in (LiCl-KCl)eut., and also to...
A pyrochemical technology for reprocessing spent nuclear fuel (SNF) and fast reactors is being implemented. One of the redistributions of pyrochemical technology is the electrochemical reduction of uranium dioxide (actinide oxides) with lithium in a LiCl - Li2O melt (1-2 wt.%) uranium dioxide and rare earth oxides at 650 °C. To test the technological regimes of the reduction process, we used a...
In Purex process, Ruthenium is one of the troublesome fission products due to its complex chemistry and presence of multiple oxidation states &some extractable stable complexes in nitric acid medium. The higher concentrations of both stable and radioactive ruthenium isotopes, pose many challenges. During the reprocessing of FBR spent fuel, the tri and tetra nitrato complexes of ruthenium get...
New nuclear fuel cycles include reducing the long-term radiotoxicity of nuclear waste by separation and transmutation of long-lived transplutonium elements. Therefore, selective recovery of transuranic elements, especially actinides (III) – americium and curium – from high-level waste generated during spent nuclear fuel reprocessing is an important issue. Processes for extracting americium...
The present study demonstrates the use of inorganic ion-exchanger (IX) to condition the high-level liquid waste (HLW) by selective separation of one of the major radionuclide, cesium-137 (137Cs) from it. 137Cs possesses a broad range of potential applications in societal and agricultural area. In addition to this, the selective separation of 137Cs from HLW would drastically bring down...
In recent years, several countries, including Russia, have been developing a pyrochemical (anhydrous) method for spent nuclear fuel (SNF) reprocessing. Molten salts have several advantages, such as thermal and radiation stability, a wide electrochemical window, etc. They can be used practically at all technological stages of SNF processing.
The first stage of pyrochemical reprocessing...
Long-lived minor actinides (MA) like, Neptunium, Americium, and Curium are the major burden of nuclear power. Long-lived MAs have not yet been used as nuclear fuel. Therefore, the transmutation of long-lived MAs is introduced as an alternative to direct final disposal. In current work, we compare the performance of MA transmutation in a critical Single-fluid Double-zone Thorium-based Molten...
As a result of the operation of nuclear reactors, a certain amount of Cm is produced, which is included in the minor actinides series (MA). Among the long-lived Cm isotopes, Cm243 and Cm245 should be noted. Their fission cross section is over 2.5 barn. In this regard, Cm can be used as a fuel in a fast neutron nuclear reactor.
For the scientific research, was used a model of the RBEC reactor...
Due to the complexity of the fast reactor project and its technical uncertainty, the design needs a long period. In order to improve design and research ability of fast reactor and develop the technology of digital reactor, bring in the model-based systems engineering method for requirement analysis, function decomposition and architecture design and weigh the overall design, to the benefit of...
Molten Salt Reactors, as a whole reactor category, belong to the GenIV reactors. They can be designed as thermal, epithermal of fast systems for variety of applications. Especially the Molten Chloride Fast Reactors (MCFRs) provide very hard neutron spectra and very high neutron economy. Hence, MCFRs can be operated as breeders in the closed U-Pu and Th-U cycles or as breed-and-burn reactors in...
During the operation of the BN-800 reactor, a large amount of experimental data has been accumulated on critical states, the effectiveness of the control system, etc. It should be noted that the loading of the hybrid core in the initial period was constantly changing: different ratios of fuel assemblies with uranium fuel and MOX fuel, as well as the number of fuel assemblies in the...
The current energy production, resulting from the concepts related to the fission of fissile nuclides, nuclear energy, is of the order of 397,650 MWe produced by the 449 nuclear plants in operation and another 54,364 MWe to be supplied by another 54 under construction on the planet [2 ], data that demonstrate the growth in installed capacity and the installation of electrical energy from...
The paper presents the results of the end-to-end mathematical modeling of the BN reactor with integral equipment layout. The developed approach permits to validate RP characteristics and to study the process of the transfer of the predecessor of the delayed neutrons with the primary circuit coolant in the conditions of stratified current.
The approach includes a complex of specially developed...
The analysis and elaboration of the thermal-hydraulic characteristics based on the results the reactor plant (RP) commissioning allow validating the algorithms of passing the modes and sufficiency of the margins applied in the project related to the thermal-hydraulic characteristics of the main equipment.
The paper presents the comparative analysis of the start-up algorithms and operation...
ENEA has a long-lasting expertise in the design of Gen IV nuclear reactors, in particular the ones cooled by liquid Lead (LFRs). In the EU context, through the participation to the FALCON Consortium, ENEA is pursuing all the activities required to support the construction of ALFRED – the European demonstrator of the LFRs – in Romania.
S/U analyses are a paramount step for the licensing of...
A comprehensive CFD model of reactor pool of liquid sodium cooled pool type 600 MWe fast reactor design along with immersed reactor components is developed for detailed thermal hydraulic studies. Hot and cold pools along with immersed components represent the primary heat transport system. The two pools are physically separated by inner vessel, which completely envelopes the hot pool. Cold...
From a neutronic point of view, the effects of thermal expansion on the reactivity of a reactor core are an important feedback mechanism, both in steady-state and during many postulated accidents sequences. It is therefore necessary to model the expanded configuration in terms of shapes, densities and volumes as accurately as possible. Unfortunately, this is not easy for those regions that...
Fast reactor (FR) operation in closed nuclear fuel cycle (CNFC) is accompanied by the change in isotopic composition of a recycled fuel during a prolonged period of time (10-30 years). Series of similar calculations are required to determine optimal parameters of core charge and operational conditions during this transient phase of FR operation. To solve this problem, a RTM software complex...
The tally cards F6 and F7 in MCNP program allow users to calculate reactor power. After a time of operation, the fission products increased, which caused the delayed energy in the reactor. Thus, the power directly calculated by F6 and F7 would not correspond with the real value, and for the fast reactor, the energy distribution of fuel and other structural materials will also deviate from the...
For power production and 233U breeding from thorium, a preliminary neutronic design of an Accelerator-Driven Sub-critical System (ADS) is presented. The ADS reactor core design with “HEU–Thorium Oxide fuel” was coupled with proton accelerator and spallation target. The neutron source (ADS system) feasibility of HEU burning and isotopes production was evaluated. The multiplication factor Keff,...
The isotopic kinetics code BPSD is developed by IBRAE RAN in “Codes of New Generation” subproject of “Proryv” project. BPSD solves fuel, absorber (boron carbide, dysprosium hafnate) transmutation, coolant (lead, sodium) and steel activation problems. Moreover, it carries out activation and residual heat calculations of materials. BPSD is intended to model materials, applied in fast reactors...
The computational code COMPLEX for radiation safety assessment of reactor and nuclear fuel cycle facilities is a set of programs (modules) combined by exchange data files and a pre- and post-processing system. The code is being developed in the "Codes of new generation" subproject of the "Proryv" project. The code includes the following modules:
- reactor core calculation modules based on...
High level liquid waste (HLLW) is often stored in large capacity horizontal cylindrical tanks especially in fast reactor fuel reprocessing plants. However, these huge tanks when partially filled, pose safety concerns due to seismicity. Violent sloshing during an earthquake-induced Fluid-Structure Interaction (FSI) can lead to catastrophic effects such as structural failures, gas entrainment...
The uncertainities of evaluated nuclear data represent one of the most important sources of uncertainity in the reactor physics simulation. The improvement of these data used is requiered for the development, safety assesment and licensing process of a reactor. Is generally recognised the need for further investigation (experimental included) regarding the uncertainities on some main...
This paper proposes a more stringent method for customizing project rules. This method customizes the comprehensive rules of the project and component reference database on the digital plant design platform based on some general design codes, standards and item classification principles in nuclear engineering, digitalization requirements in reactor design, plant layout, project management,...
Calculations of non-stationary processes of fast neutron reactors taking into account the spatiotemporal dependence of the neutron field is a rather complex process due to the significant influence on the calculation results of delayed neutrons, which make up a very small, less than a percent, part of all the neutrons of the reactor in its critical state. This circumstance is due to the fact...
The advanced development of NPE assumes gradual embedding fast neutron reactors, which enshures the most complete use of the uranium and thorium resources. Even on the theoretical level there are some alternative solutions for organization of operation of reactor facility and the fuel cycles. The experimental verification requirs the considerable amount of time and serious material resources....
To date spent nuclear fuel (SNF) reprocessing is a promising field of study. More than 370 thousand tons of SNF have been accumulated in the world and 10-12 thousand tons are added to this amount annually. In Russian Federation, ~25000 tons of SNF were accumulated according to the data of 2018. Pyrochemical technology of SNF processing, which is supposed to substitute aqueous technologies, is...
The National Institute for Nuclear Research (ININ) of Mexico, participates in the IAEA-CRP on Neutronics Benchmark of the Chinese Experimental Fast Reactor (CEFR) Start-Up Tests, which was proposed by China Institute of Atomic Energy (CIAE). The Mexican participation in this Benchmark is focused in two main goals: the first one, the use of SERPENT code for the generation of reference solutions...
The report presents the results of comparing the calculated data and readings of devices for monitoring water leakage into sodium, observed during a real leak in the BN-600 steam generator.
BN-600 implemented a section-modular scheme of a sodium-water steam generator. The damage of the heat exchange surface of the BN-600 steam generator occurred mainly in the initial period of plant operation...
CFD modeling was extensively used for the development of high-temperature furnaces for the carbothermal synthesis of uranium and plutonium nitrides and a furnace for the sintering of mixed nitride uranium-plutonium fuel pellets. This equipment is intended for use at the Pilot Demonstration Energy Complex (PDEC) being constructed in Seversk, Russia. The СFD-model of the carbothermal synthesis...
