The helium cooled high-temperature fast-spectrum reactor (GFR) with closed fuel cycle is one of the six GEN IV reactors selected by the Generation IV International Forum (GIF) to be developed for the foreseeable future. The European reference concept of the GFR technology is a unit with an envisaged power of 2400 MWth, which is currently in the pre-conceptual design phase. Prior to the...
The elements of the fuel assembly including individual fuel pins are affected by the coolant flow. This can lead to mechanical vibrations. The cyclic loading of the fuel element cladding material accompanying these vibrations causes an additional effect on the fuel element material. This can cause the damage of the fuel cladding material, especially in its contact with the spacer elements. The...
In the context of Lead-cooled Fast Reactor development and safety assessment, the flow blockage in a fuel sub-assembly is considered among the most relevant issues to be addressed. Hence, the event shall be postulated assessing its consequences, also considering that grid-spaced fuel assemblies could partially mitigate the occurrence of sudden blockages with respect to wire-spaced fuel...
In this work, several CFD simulations of the hexagonal 61-pin fuel bundle replica of the Thermohydraulic Research Lab at Texas A&M University were performed. This fuel assembly geometry is a helically wire wrapped bundle of rod pitch-to-diameter ratio of 1.89 and helix pitch-to-diameter ratio of 29.93, as in Sodium-cooled fast breeder reactors (FBR). The experimental activity has produced a...
The system thermohydraulic code HYDRA-IBRAE/LM is designed for the simulation of non-stationary thermohydraulic processes in liquid metal and water circuits of fast reactors under normal operating conditions, anticipated operational occurrences and accidents. The code uses a two-fluid model in all flow regimes except for dispersed annular flow, where a three-fluid model is applied. Besides...
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...
Since 2010, domestically produced software required for design decision making and safety assessment of nuclear power plants with fast reactors has been developed in Russian Federation under “Proryv” project, namely, under one of its subprojects – Codes of New Generation.
As a priority, a task was set and successfully accomplished covering the development of 24 software products addressing...
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...
The cover gas region of sodium cooled fast reactors is always being subjected to intense ionization radiation field apart from radioactive aerosols and gases. The radiation produces significant ionization of the medium resulting in large amount of bi-polar ions. The acquisition of electrical charge by sodium aerosols in cover region under bipolar ionic atmosphere draws special attention as it...
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...
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...
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...
There remain challenges in studies of properties and irradiation behaviors of mixed oxide (MOX) fuels, which aims at reduction in volume and toxicity of high-level radioactive wastes, because of the influential factors such that the fuel reaches very high temperature exceeding 2000 K and oxygen content in the fuel continuously varies depending on surrounding conditions. High temperature and...
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...
Liquid metal fast reactor have a prominent role in the roadmap of the Dutch nuclear stakeholders. As nuclear service provider in the Netherlands, the Nuclear Research and consultancy Group (NRG) has established an elaborate program on liquid metal thermal hydraulics. This paper describes the thermal hydraulic design and safety support activities of NRG. The paper will start with the...
On the basis of detailed review, the fuel types were proposed for the new design of the ALLEGRO gas-cooled fast reactor. The first core will be built with MOX or UOX fuel in 15-15Ti stainless steel cladding. These fuel types have been widely used in different sodium-cooled fast reactors. The second core of ALLEGRO will use refractory fuel. The primary candidate is carbide fuel – (UPu)C or UC...
In order to substantiate the design characteristics of the steam generator of the BREST-OD-300 reactor plant (RP) developed at NIKIET JSC, IPPE JSC carried out thermohydraulic tests of various models of the lead-heated steam generator. Initially, to confirm the design characteristics and thermal-hydraulic stability at the parameters of the nominal, partial and start-up modes, a model of a...
In India, a structured R&D program on the development of metallic fuel and associated fuel cycle for Fast Breeder Reactors (FBRs) is undertaken so as to realize commercial metal fuel FBRs in the future. Towards this, initially test irradiation of sodium bonded metal fuel pins in Fast Breeder Test Reactor (FBTR) core was proposed and hence the pin design for various compositions of metal fuel...
The higher level of modeling for dynamic liquid metal boiling is important for comprehensive analysis of neutron-physical and thermohydraulic characteristics of fast reactors cores for safety justification at accidental conditions (UTOP, ULOF).
Experimental data obtained at IPPE JSC have showed that boiling of liquid metals in fuel assemblies of fast reactors has a complex structure which is...
This paper describes the verification work of SARAX code for the transient analysis of a sodium-cooled fast reactor (SFR). The Advanced Burner Test Reactor (ABTR) benchmark created by Argonne National Laboratory (ANL) was modeled and calculated. The reference core is the 250 MWt sodium-cooled fast reactor, which includes neutronics calculation of the core at the beginning of equilibrium cycle,...
A high-density annular MOX fuel pellet fabrication technology has been developed for producing a low O/M ratio of less than 1.97 for fast reactors. The low O/M ratio sintered pellets aim to suppress the fuel-cladding chemical interaction (FCCI) at high burnup, and a simplified MOX pellet fabrication process (Short process) is a new production technology for this MOX fuel. The short process is...
In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems (DHRS) for safety enhancement against severe accidents which could lead to core melting. To clarify the natural circulation phenomena in a reactor vessel during operation of a decay heat removal system, water experiments have been conducted using a 1:10 scale experimental...
Before the definitive shutdown of the Phénix reactor, a series of end of life tests were performed in 2009 and 2010, by CEA (Commissariat à l’Energie Atomique et aux Energies Alternatives), EDF (Électricité de France) and AREVA. The main objectives were to enlarge experimental database for the research and design of Sodium cooled Fast Reactors (SFR). Due to this important opportunity, the IAEA...
Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) is a joint project between the U.S. Idaho National Laboratory (INL) and the Japanese Atomic Energy Agency (JAEA) to investigate the transient fuel performance of metallic and MOX fuels. The project has the specific goals of experimentally evaluating the transient failure modes of high burnup metallic and MOX fuels, guided by...
As a possible part of the safety instrumentation in liquid metal cooled fast reactors, the Eddy Current Flow Meter (ECFM) is an important and robust tool for continuously monitoring the coolant flow and detecting coolant blockages in the reactor subassemblies for coolant temperatures of up to 700 °C. This inductive sensor can be placed directly above the subassemblies, where changes of the...
The Fast Flux Test Facility (FFTF) at the Hanford site in Washington was a 400 MW thermal, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. FFTF operated from 1980 until 1992, providing the U.S. Department of Energy (DOE) with the means to test fuels, materials, and other components in a fast...
In IBRAE RAN in “Codes of New Generation” subproject of “Proryv” project one- and two-phase models are being developed to simulate heat and mass transfer processes in the separate elements of nuclear reactor. Those models are realized in the LES and DNS CONV-3D code.
The one-phase models are based on the algorithms with small scheme diffusion, for which the discrete approximations are...
In the framework of the European Commission call for proposal “Horizon 2020”, the project Plutonium Management for More Agility, called PuMMA, is granted. This project starts in October 2020 and will last four years. A work package is dedicated to the behaviour and safety of mixed oxide fuels with high plutonium content, which is essential for plutonium multi-recycling or plutonium burning in...
In the design study of sodium-cooled fast reactor, various activities from sensitivity analysis on whole plant dynamics using simple model to detailed analysis on local phenomena of interest are being performed. In conventional way, the analyses on whole plant dynamics and local phenomena are performed individually and the mutual interaction between them are considered through the settings of...
The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, is a loop type, sodium cooled fast reactor. Its main aim is to provide experience in fast reactor operation, large scale sodium handling and to serve as a test bed for irradiation of fast reactor fuels & materials. India has been operating FBTR with Mixed Carbide Fuel as the driver fuel since...
In the framework of reactor safety analysis of Sodium cooled Fast Reactors (SFR) using applicable code systems, the CEA and JAEA are involved in the achievement of SIMMER-V (owned by JAEA and co-developed by JAEA and CEA) developments dedicated to the simulation of the Severe Accident (SA) events of SFR.
The demand for new physical models into SIMMER-V rises in SIMMER previous versions...
The thermogravimetric method was used to study the behavior of uranium nitride and mixed uranium-plutonium nitride (MNIT) in a helium flow and a helium with nitrogen gas mixture at temperatures up to 2173 K. When heated in helium in the low-temperature range (˂1773 K), a mass loss was found, which amounts to hundredths of a percent. In this case, mass loss occurs in 2 stages, accompanied by...
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...
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...
The most dangerous of the minor actinides is americium. Transmutation of external americium in the fuel of a fast reactor is possible when its content is over than 1% heavy atoms, however the lower content of an americium, on the contrary, it will accumulate. But curium isotopes with a high heat release are formed from it, complicating the unloading of spent assemblies. Therefore, the content...
For the modeling of severe accidents in a sodium-cooled fast reactor coupled multiphysics EUCLID/V2 code is being developed in Russian Federation in Codes of New Generation subproject of “Proryv” project. Multiphysics code allow calculating all relevant processes occurring during severe accident: reactor power change including due to boiling and melting, coolant boiling and dryout, cladding...
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...
The Strategy is based on the formation of a two-component nuclear energy based on a closed nuclear fuel cycle with fast neutron reactors.
The solution of long-term tasks of creating a two-component nuclear power with a closed NFC based on fast reactors is associated with the need to create a system for training, attracting and developing young professionals based on new...
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...
The United Kingdom was one of the countries pioneering fast reactor development and demonstration with the construction and operation of two fast neutron spectrum reactors at the Dounreay site; multiple support facilities including zero-power and thermal hydraulic facilities; and demonstrated several sustainable and closed fuel cycles. Since the end of the UK Fast Reactor Programme in 1994...
Analytical evaluation of severe accidents (SAs) in sodium-cooled fast reactors (SFRs) becomes increasingly important. The progress of the SAs has been previously evaluated by transferring the analytical results between the multiple analysis codes with different roles. In this study, a new code named SPECTRA (Severe-accident PhEnomenological computational Code for TRansient Assessment) was...
The programme described here is part of a UK government investment in Nuclear to be delivered by a collaboration of UK Government, The UK National Nuclear Laboratory (NNL), Industry and Academia. The programme will contribute to international understanding of development and demonstration requirements for pyro-processing as well as emerging applications for salt separations enabling a broad...
This contribution presents an overview of models of an aerosol module designed to simulate the behavior of fission products in the circuits and compartments of nuclear power units with fast reactors with sodium or lead coolants. Aerosol module AEROSOL-LM is included in the thermal-hydraulic HYDRA-IBRAE/LM code. Together they represent a unified code with a common interface for calculating the...
Dense nuclear fuel for fast reactors (FR) is the preferred option. In the Russia, as part of the “PRORYV" project, the development of key technologies of closed nuclear fuel cycle (CNFC) for FR with dense mixed nitride uranium-plutonium fuel (MNUP) is underway. MNUP is a new complex product in the field of nuclear power technologies. CNFC with FR ensures:
− no spent nuclear fuel (SNF)...
Collaboration and support among national laboratories, industry, universities, and research and development organizations are vital to not only maintain a skilled and competent nuclear workforce but also to avert the risk of human resource shortages. However, despite numerous efforts in coordinating and promoting nuclear education, there is still a lot to be done for developed and developing...
The structure of pool type sodium-cooled fast reactor (SFR) is complex, which leads to the complicated thermal-hydraulic phenomena in the process of natural circulation for decay heat removal. The determination of natural circulation flow path and the decay heat removal capacity of natural circulation of each flow path are issues to be considered in the design of SFR. Core flow distribution,...
Transmutation of minor actinides (MA) into stable or short-lived ones by their irradiation in reactors will alleviate the problem of long-term activity of spent nuclear fuel (SNF), increase the efficiency of nuclear fuel due to energy produced by MA fission, and also accumulate and produce useful radionuclides. The economic efficiency of closed-cycle nuclear power cannot be achieved without MA...
Obninsk Institute for Nuclear Power Engineering has been realizing education and training of specialists for nuclear industry since establishing in 1950s. Educational programs are developed for nuclear power plants as well as for research and development institutions of Russian Federation and abroad. Due to close connections with scientific Institute of Physics and Power Engineering named A.I....
The processes developed for partitioning of trivalent actinides (TA) from high-level liquid waste (HLLW) generated during reprocessing are all focused on single-cycle approaches for waste minimization. In this process the formation of third phase has to be avoided. Hence, a phase modifier is often employed in most of the processes in vouge, even though the use of the later is more desirable....
ALFRED is the demonstrator of Lead Fast Reactor (LFR) technology. According to strategic documents (at national level and of FALCON international consortium), it is planned to be built on Mioveni nuclear platform. An experimental infrastructure consisting of six experimental facilities (ATHENA, HELENA2, ELF, ChemLab, HandsON, Meltin’Pot) and a coordination Hub is planned to be built on the...
The EUCLID/V2 integral multiphysics computer code is designed for the safety analysis and justification of the new generation NPPs with liquid metal cooled fast reactors under normal operating conditions, anticipated operational occurrences, design basis accidents and severe accidents. The EUCLID/V2 code includes the system thermohydraulics module (HYDRA-IBRAE/LM), spatial time-dependent...
This paper presents a comparison of the efficiency of minor actinides (MA) burning in various type of nuclear reactors with a fast neutron spectrum. A set of criteria for comprehensive comparison of reactor technologies, based on the INPRO/IAEA KIND approach to multi-criteria assessment, has been prepared. This set of criteria includes indicators in such areas as the efficiency of MA burning,...
The International Atomic Energy Agency (IAEA) supplements and supports nuclear research and development with many efforts to improve and make nuclear data more accessible. Through technical community building research projects, and tool development, the Nuclear Power Technology Development section (NPTDS) provides many services and opportunities to amplify the research of member states’...
The transition to a two-component nuclear power structure using thermal (TR) and fast reactors (FR), as asserted by the «Russian nuclear power development strategy to 2050 and outlook to 2100» (Strategy-2018), is directed at finding optimal solutions and resolving relevant issues pertaining to the currently established nuclear energy system in Russia. A core issue in this regard is managing...
Alkaline halide melts and alkaline earth metals are used for the electrochemical reduction of metal oxides to their metallic forms. In practice, fluoride, chloride, and mixed chloride-fluoride melts of alkali and alkaline earth metals are used most often. Graphite is usually applied as an inert anode material in these media. However, during the electrolysis of oxide-halide melts, carbon is not...
Integral computer codes SOCRAT-BN have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) in the frame of the Federal Target Program «New-Generation Nuclear Power Technologies for the Period 2010–2015 and up to 2020». The first version SOCRAT-BN/V1 was developed for the period 2010-2014 to simulate design basis (DBA) and beyond design basis accident...
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 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...
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...
Artificial Neural Networks (ANN) are presented as a very powerful tool for modelling complex systems. This approach is becoming increasingly widespread and it has a great potential for nuclear reactor applications. In this work, an ANN is developed for predicting sodium void effects in a large Sodium Fast Reactor core and their spatial interrelations.
The ultimate goal is to provide more...
While developing an advanced BN reactor plant the tasks were put to reduce a reactor plant cost with obligate meeting of safety requirements and reliability increase of reactor plant equipment and systems.
A purification system with cold traps (CT) placed in the reactor vessel (built-in purification system) was applied in the large BN reactor plant for primary sodium purification.
The...
A large 3600 MWth European Sodium Fast Reactor (ESFR) design was proposed in 2000s. It is studied now in the EURATOM ESFR-SMART project. A new core configuration with several new safety measures, including a reduced to a near-zero value coolant void reactivity effect, mainly due to introduction of a sodium plenum above the core and core flattening, has been established recently. We investigate...
Large and Small Rotatable Plugs (LRP & SRP) form part of top shield and are used to position transfer arm over any required subassembly location during fuel handling. The annular gap between the Roof Slab and LRP as well as between LRP and SRP is sealed with the help of two types of elastomeric seals - Primary inflatable seals and secondary back up seal. These seals have a design life of 10...
The CEA, together with the NNC, has carried out a feasibility study with regard to conducting an in-pile test program - the future SAIGA program (Severe Accident In-pile experiments for Gen-IV reactors and the Astrid prototype) - on the degradation of a SFR fuel bundle with molten fuel discharge device which are planned to be housed in the IGR reactor (Impulse Graphite Reactor operated by...
To achieve in-vessel retention for mitigating the consequences of core disruptive accidents (CDAs) of sodium-cooled fast reactors, controlled material relocation (CMR) has been proposed as an effective safety concept. CMR is not only aiming at eliminating the potential for exceeding prompt criticality events that affect the integrity of the reactor vessel, but also enhancing the potential for...
Concept of the BREST reactor with lead coolant and dense heat-conductive nitride fuel envisages the development of an equilibrium core with complete breeding of fissionable nuclides in the core (core breeding ratio of ~ 1) without a blanket compensating for reactivity reduction due to fuel burn-up and fission product build-up. This makes it possible to operate the reactor in the period between...
Safety studies of fast reactors are carried out on a medium sized core and found that, under Unprotected Transient Over Power Accidents (UTOPA) there is fuel melting and there is feedback due to in pin fuel motion. From the UTOPA analyses it is found that, the in pin fuel motion feedback reduces the peak reactor power and hence it reduces the hot spot clad and coolant temperatures at the end...
Fast Breeder Reactor -1&2 (FBR-1&2) is a sodium cooled, pool type, Mixed Oxide (MOX) fuelled reactor with two sodium loops (primary and secondary). The design of this reactor is based on experience from Fast Breeder Test Reactor (FBTR) and prototype Fast Breeder Reactor (PFBR). Decay Heat Removal (DHR) system removes decay heat from the reactor after shutdown to ensure adequate cooling of core...
Energy decarbonisation, through the transition from fossil fuels to V-RES electricity production and the electrification of transport & heating sectors, may jeopardise the electricity supply security on the long term, because of the growing power demand and the increased production volatility. While advanced and modular reactor designs can make nuclear an attractive low-carbon solution to...
In order to enhance the competitiveness and to reduce the construction cost of the future industrial Sodium Fast Reactors (SFR), several options are explored which need further R&D studies or design assessment. Among them, the possibility to reduce the size of the reactor vessel has been investigated through the reduction of the core diameter and the increase of the power density thanks to...
An In-vessel core catcher located in European type SFR is a safety design feature to guarantee the integrity of the SFR during a core-melting accident. The core catcher collects and distributes the relocated melt from the core region via discharging tubes to avoid firstly a significant molten pool in the core, and secondly the local thermal attack on bottom part of the vessel. The heat...
The energy trilemma and UN-SDG 7 are drivers for energy research to support the UK governments net-zero emissions law. Nuclear reactors are a highly attractive candidate for reliable, 24/7 available, low-carbon electricity generation. However, current technology reactors and their related fuel cycles suffer from unreasonably high cost, a lack of sustainability, and a waste problem due to 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 identification of the R&D priorities and needs is a necessary step to complete the development, up to the qualification and demonstration, of the solutions envisaged for LFR technology. In particular, this is of paramount importance to allow the design, licensing and construction of industrial systems.
In the present work, starting from the key scientific aspects (including lead chemistry...
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...
The purpose of this article is to investigate the use of fast reactors for changing the isotopic composition of Pu for a better reuse in thermal reactors. The possibility to change or adjust (the term «improve» can also be used) the isotopic composition of Pu from MOX SNF of thermal reactors is determined by the fundamental characteristic of fast reactors – their capability for nuclear...
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,...
The OpenFOAM Fuel BEhavior Analysis Tool, i.e. OFFBEAT, is a research-oriented multi-dimensional fuel behavior solver under development at the Laboratory for Reactor Physics and Systems Behaviour at the EPFL and at the Paul Scherrer Institut, in Switzerland. OFFBEAT relies upon the C++ library OpenFOAM and it aims to be a complement to traditional fuel performance codes for the study of 2D and...
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...
The OSCAR-Na code has been developed during the last decade to calculate the mass transfer of corrosion products and related contamination in the primary circuit of sodium fast reactors (SFR). Indeed, even if fuel cladding corrosion appears to be very limited, the contamination of the reactor components plays an important role in defining the design, the maintenance and the decommissioning...
A long-term objective of the Advanced Fuels Campaign (AFC) is the investigation into enabling technologies that allow for improving nuclear fuel performance and for the transmutation of minor actinides in sodium fast reactor. As part of this development, candidate fuel compositions and forms are irradiated in a cadmium-shrouded positions at the INL’s Advanced Test Reactor (ATR), and they are...
Lead fast reactors are of particular interest thanks to the characteristics of the coolant which simplifies numerous plant choices while maintaining high levels of safety and reliability. In Europe, ALFRED is the main technology demonstrator and leveraging on the knowledge developed in the evolutionary generation III+ plants it makes use of passive safety systems, for example for the decay...
Control Rod Drive Mechanisms (CRDM) along with their control rods are used for control and safe shut down of Fast Breeder Test Reactor (FBTR). Lower part of CRDMs which is partially immersed in sodium and nested ripple type welded disc bellows are used to prevent entry of sodium to the annular spaces between the concentric tubes in the lower part. Translation bellows are used to prevent the...
The Molten Salt Reactor (MSR) is unique, with respect to other Gen IV concepts as well as current LWRs, in the fact that the liquid fuel comes with a slew of safety-relevant features, which are chemically distinct from those otherwise encountered. These features are often neglected in the scope of neutronics-based investigations into the topic, where more heed is paid to the isotopic...
Today the investigations have been completed on helium-bonded fuel pins with mixed uranium-plutonium nitride of BN-600/BN-800, BN-1200 and BREST reactors types after irradiation as a part of ten EFAs of BN-600 reactor up to maximum burn-up of 7.5;6.0 and 4.5at%, respectively. Also the PIE of mixed nitride pins of BREST type with helium and lead sub-layers after intermediate tests to maximum...
This paper reviews the status of fast reactor fuels. The main focus of the development of fast reactor fuel is their potential for actinide transmutation and high burn up. Metallic, oxide, carbide, nitride and dispersion type fuels are being used as fast reactor fuels. Metallic fuels, because they produce an extremely hard neutron spectrum, are neurotically ideal for fast reactors. As compared...
In order to protect the key equipment from high temperature in fast reactor, main pumps and main vessel is shielded by single or multiple hot screens, forming narrow fluid gaps. However, these fluid gaps bring some difficulties in seismic analysis by introducing the fluid structure interaction effect. Added mass, a simplified but important parameter of fluid structure interaction effect, is...
Current methods for the design of high temperature fast reactor components are deterministic, often based on deterministic structural analysis compared to factored design material data. These methods often produce very conservative designs and the exact design margin -- for example, the probability of premature component failure -- often cannot be easily quantified. A statistical design method...
To enhance safety, an emergency heat removal system (EHRS) is provided as a part of the reactor plant. One of the main elements of this system is an air heat exchanger (AHX), equipped with a device for air flowrate control with passive opening principle, which is a gate.
To ensure operability of this gate, high-temperature experimental test facility was developed and manufactured. This test...
Uranium-zirconium carbonitride has been developed at the LUCH FSUE and is a high-density high-temperature fuel with high heat conductivity capable of being used in various types of reactors, including fast reactors. The main problem hindering wide application of this fuel is insufficient knowledge of its behavior under irradiation, especially at high burnup. In the USSR, HEU UZrCN (96% by...
In the framework of the current implementing arrangement on France-Japan collaboration on Sodium-cooled Fast Reactors (SFRs) from 2020 to 2024, the R&D tasks called “Thermodynamic and Kinetic Studies of Core Material Mixture” is intended to improve models on material interactions at thermodynamic equilibrium and kinetics of reactions for use in severe accident simulation codes with...
At the moment, in the world nuclear power industry there are proposals for the transition from oxide to other types of fuel, which can be much more cost-effective and used in the technology of closing the nuclear fuel cycle based on fast reactors. One of such fuels is uranium-plutonium nitrides. During the fission of uranium and plutonium nuclei in fast neutron reactors, a large number of...
The BERKUT-U mechanistic fuel performance code has been designed at Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) since 2012 in frame of “Codes of new Generation” subproject of “Proryv” project. The code is intended for self-consistent computational simulation of the stress-strain state and temperature distribution in fuel rods with nitride or oxide fuel, with a gas...
In a pool type sodium cooled fast reactor, in case of detection of failure of a fuel subassembly (FSA) by global delayed neutron detection system, localization of failed subassembly would be done using the Failed Fuel Location Module (FFLM). This is achieved by sampling sodium at exit of each subassembly and looking for presence of delayed neutrons. For a 500 MWe prototype design, as part of...
The Versatile Test Reactor (VTR) is a proposed fast neutron spectrum test facility that will provide irradiation capabilities not currently available within the U.S. The Idaho National Laboratory (INL), in conjunction with five other U.S. national laboratories, 19 universities and 10 industry partners, is working to develop the VTR to provide an irradiation-testing facility capable of...
The Fast Reactor concept has been proposed by Generation-IV initiative as a potential candidate to develop safe, sustainable, reliable, proliferation-resistant and economic nuclear energy systems (GIF, 2002). Within fast reactor core, fission chain reaction is sustained by fast neutrons which result in a much higher and harder neutron flux than that of thermal reactors. This high neutron flux...
The next generation of nuclear energy systems, also known as Generation-IV reactors are being developed to meet the highest targets of safety and reliability, sustainability, economics, proliferation resistance and physical protection, with improved performances with respect to plants currently operating or presently being built. Among the proposed technologies, Lead-cooled Fast Reactors...
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...
One of the most important problems in the further development of nuclear energy, from the point of view of its public acceptance, is the problem of safety. Thus, the development of new concepts for nuclear fission reactors with so-called "intrinsic safety" is a very urgent task. An equally important problem for the sustainable development of nuclear power is the need to expand the fuel base by...
