Towards the materials development for future cores of sodium cooled Fast Breeder Reactors of India, an improved clad material named as Indian Advanced Fast Reactor Clad (IFAC-1) was developed in the form of clad tubes (20% cold worked) of length more than 3 meters. The composition optimization of IFAC-1 SS was arrived at based on the extensive evaluation of tensile and creep properties on 15...
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...
Experience and competences obtained in the course of BN reactor technology formation from BR 5/10 to BN-600, BN-800 provide the possibility for commercial deployment of this technology in a two-component nuclear energy system.
The paper presents the results of implementation of commercial reactors BN-600, BN 800, development and substantiation of a commercial power unit with a new-generation...
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. ...
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...
The IAEA supports Member States in the area of advanced fast reactor technology development by providing a major forum for information exchange and via collaborative research programmes (CRPs). Under the broad framework of coordinated research activities in the area of fast reactors, the IAEA proposes and establishes several coordinated research projects (CRPs), aimed at improving Member State...
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...
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...
China Experimental Fast Reactor (CEFR) is a sodium-cooled fast-spectrum reactor built in China Institute of Atomic Energy, Beijing. The thermal power is 65MW. In the first cycle the reactor core was loaded with uranium dioxide fuel enriched at about 64%.
In the physical start-up tests in 2010, a series of experiments were conducted, which not only made an essential part of the reactor...
India has a well established Nuclear Power Programme with currently 22 reactors in operation and 6 under construction. The total contribution of nuclear power is 6780 MWe. Fast breeder reactors are expected to play a major role in providing energy security to the country by effective utilisation of the spent fuel from the existing reactors and by utilisation of thorium on a later date. The 40...
In the reactor plants, which have inherent safety properties, high-heat conducting nitride uranium-plutonium fuel and lead coolant are used. Nitride fuel is less ductile than oxide one, and the resource of a nitride fuel element with a helium gap is limited primarily by the touch and deformation of the shell by fuel (and not by its radiation damage). Use of the liquid-metal sublayer in the...
By introducing high stability oxide nanoparticles into FeCrAl alloy matrix, the design concept of dispersion strengthening can be realized. Oxide dispersion strengthened (ODS) FeCrAl alloy has been considered as a promising candidate for cladding materials due to their excellent mechanical strength, remarkable structural stability and chemical durability at elevated temperature. ODS FeCrAl...
This paper provides an overview of fast reactor research and development efforts in the United States. Fast reactors are envisioned for a wide variety of actinide management strategies ranging from actinide destruction in closed fuel cycles to enhanced uranium utilization. With successful technology development, fast reactors are also intended for electricity and heat production, as being...
Under the framework of coordinated research activities of the International Atomic Energy Agency (IAEA), the China Institute of Atomic Energy (CIAE) proposed a coordinated research project (CRP) to develop a benchmark based on the start-up tests of the China Experimental Fast Reactor (CEFR). 29 international organizations from 17 countries are participating in this CRP. Among the different...
The Gear Test Assembly (GTA) is an experimental apparatus designed to test mechanical components, specifically gears and bearing, used in advanced fuel handling systems of liquid-sodium cooled fast-spectrum nuclear reactors. Reviews of existing documentation indicated a lack of testing for these specific mechanical components used in the construction of advanced fuel handling systems. Most...
The China Institute of Atomic Energy (CIAE) proposed some of the China Experimental Fast Reactor (CEFR) neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activity. The coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests” was launched in 2018. This benchmark aims to perform validation and verification of the...
Austenitic stainless steel (SS) 316 L(N) is the material for Indian Fast Reactor internals such as main vessel, grid plate, core support structure, core catcher, control plug, inner vessel, IHX etc, while the safety vessel is made of SS304 L(N). These permanent components experience a temperature of 350°C-550°C and accumulate neutron doses of < 1 dpa in their life time. An irradiation...
The IAEA Coordinated Research Project (CRP) focuses on neutronics benchmark analysis of the physics start-up tests performed at the China Experimental Fast Reactor (CEFR) in 2010- 2011. CEFR is a small-size sodium-cooled fast reactor with a high neutron leakage core fuelled with uranium oxide and stainless-steel radial reflector. The primary purpose of this IAEA/CRP is to improve the Member...
Starting from 2012 the Lead-cooled Fast Reactor provisional System Steering Committee (LFR-pSSC) of the Generation IV International Forum (GIF) has developed a number of top-level strategic activities with the aim to assist and support development of Lead-cooled Fast Reactor technology in member countries and entities. The current full members of the GIF-LFR-pSSC (i.e., signatories of the GIF...
In 2018, the International Atomic Energy Agency (IAEA) established a coordinated research project (CRP) for benchmark analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. Argonne National Laboratory and Pacific Northwest National Laboratory are the lead technical coordinators for the CRP. Initiated at half power and full flow, LOFWOS Test #13 was an...
Due to its very attractive properties, Na is used for nuclear and solar applications, in similar operating ranges. It is also used to study the dynamo effect, to get a better understanding of the magnetic field of the earth, to purify metals such as tantalum or silicon for photovoltaic cells... In order to operate in reliable and safe conditions a SFR, it is necessary to master the coolant’s...
The Lead cooled Fast Reactor (LFR) has been selected by the Generation IV International Forum (GIF) as one of the most promising nuclear technologies able to meet the GIF goals and playing an increasingly important role in the international context. The ALFRED (the Advanced Lead-cooled Fast Reactor European Demonstrator) project aims to bridge the gap between the research and development...
The International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. This benchmark exercise provides the community with a valuable benchmarking opportunity for validating SFR neutronics and safety analysis tools and methods. This paper focuses on the neutronic...
The features of the processes of hydrodynamics of heat and mass transfer in cold traps are investigated. Initially, it was mainly an empirical approach that led to good results. As a result, a scientific justification was developed for the creation of cold traps of an original design for BN-350, BOR-60 and BN-600. Their capacity for impurities, one of the most important performance...
Every day at the SCK•CEN site in Mol, over 750 scientists advance nuclear research. Thanks to its state-of-the-art infrastructure and world-renowned experts, the Belgian Nuclear Research Centre plays a vital role in developing innovative nuclear applications. Its BR2 (Belgian Reactor 2) is one of the most powerful research reactors in the world, which in particular meets one-quarter of global...
Prompt detection of steam-to-sodium leaks of a sodium-heated steam generator (SHSG) is one of the important safety and economic issues to be addressed in designing and operating a sodium-cooled fast reactor (SFR). A water-to-sodium leak causes a violent exothermic sodium-water reaction (SWR) resulting in local temperature rise (>1,200C) and produces hydrogen gas and highly corrosive...
GFR is one of six GENIV reactor technologies selected as most promising by GIF, with a long history of research and development work done in various countries. Modern day GFR research is concentrated mainly in Europe, Japan and USA. European GFR reference concepts are GFR2400, a large-scale commercial reactor, and ALLEGRO, a 75 MWth technology demonstrator. The ALLEGRO development is led by...
Three Work Packages (WPs) were defined in this Coordinated Research Project (CRP) whose objective was to estimate fission-product-transportation behaviour inside the reference pool-type sodium-cooled fast reactor (SFR) volumes at different time scales under severe accident conditions.
This WP, WP-1, is defined to estimate the in-vessel source term using improved models and tools for the...
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...
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...
The objective of this work is to model a global nuclear energy deployment with fast and thermal reactors in a partially closed fuel cycle and find the structure of the global NES that is optimal from the point of economic criteria. The tool for the NES simulation used in the study is the software package MESSAGE, distributed by the IAEA. MESSAGE is a universal tool of energy planning designed...
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...
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...
Prototype Fast Breeder Reactor is a 500MWe nuclear reactor under commissioning at Kalpakkam, India. During the commissioning stage, visual inspection of the core internals to ascertain that the assembly of components inside the reactor does not impair the structural integrity of the reactor core and to scan for any foreign body debris in the core before starting up of the reactor. The Reactor...
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...
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...
Preheating of Reactor Assembly (RA) internals in Prototype Fast Breeder Reactor (PFBR) is one of the important milestones in the commissioning activity of the project. The reactor assembly internals have to be preheated using hot nitrogen to a minimum temperature of 423 0K (150°C) before initial sodium filling in the primary sodium circuit, to avoid solidification of sodium and also avoid...
Long-term sustainable development of nuclear power in Russia and worldwide is possible only if NPPs are competitive in comparison with other types of generation. At the same time, it is crucial to demonstrate a cost-effective solution to the systemic problems of the industry, including the issue of accumulating the SNF.
Modern NPPs with thermal reactors operated in an open NFC, given the...
In Prototype Fast Breeder Reactor, Secondary Sodium circuits transfer the heat from primary sodium system in the pool through intermediate heat exchangers to the steam water system for the production of steam & thereby electrical energy. These two circuits are located in Steam Generator (SG) buildings and are engineered to avoid direct ingress of hydrogenous material to primary sodium system,...
The Generation IV International Forum (GIF) Lead Fast Reactor (LFR) provisional System Steering Committee (pSSC) has assessed, in collaboration with the GIF Risk and Safety Working Group (RSWG), the safety characteristics of reference Generation IV LFR systems. The objective has been to review and identify the main safety advantages and potential challenges of the technology, to assess the...
The paper aims to present a study conducted by the engineering departments of EDF, Framatome and CEA in order to define a functional description and a sketch of the commercial industrial French Sodium Fast Reactor 1000 eMW.
This pre-conceptual design study has been carried out with the benefits of ASTRID project, which conducted the basic design of an industrial Generation IV SFR...
Prototype Fast Breeder Reactor (PFBR) has 8 Steam Generators (SG). Pre-Service Inspection (PSI) and periodic In-Service Inspection (ISI) of the SG tubes are required as a part of safety and reduce cost by increasing the plant availability. The tube wall thickness being the only barrier between sodium on the shell side and water/steam on the tube side, it is critical to ascertain the...
Following the significant scientific results and knowledge acquired from the ASTRID collaboration, the French and Japanese partners launched a new consolidation phase on the technical knowledge on sodium-cooled fast reactors (SFR) and R&D basis to maintain skills and further develop advanced technologies. After signature of an Implementing Arrangement covering the period 2020-2024 (IA2020),...
At the present stage of development in the nuclear world community, there has actually been a consensus on the dominant role of fast reactors in the future large-scale nuclear power. In addition to the main role of these reactors in supplying nuclear power with fuel in the context of an imminent shortage of natural uranium resources, as was considered in the past, nowadays an important aspect...
ABSTRACT
Safety and economy of nuclear power reactor is related to the sustainable development of nuclear energy closely, and people pay more attention to it. Small nuclear power reactors are favored by authorities and investors due to the enormous potential on small size, safety and economy. This study summarized the technical route, government support, investor interest, investigated the...
The ARC-100 is an innovative 100 MWe sodium cooled fast reactor design being developed by Advanced Reactor Concepts (ARC), LLC. The reactor employs a long-lived metallic-fueled core with low burnup reactivity swing, which facilitates citing in remote locations and/or small grids with minimal reactor maintenance during its lifetime. Argonne National Laboratory has been tasked with the safety...
Introduction
A large quantity of sodium is contained in fast breeder reactor and its neutralization is a major step of D&D of this type of reactor. Indeed, sodium treatment greatly reduces operating costs, eliminates chemical hazards and reduces radiological hazards associated with tritium.
Sodium elimination step is based on 3 types of activities:
• Draining
• Bulk sodium Treatment...
After official notification to the Romanian regulatory authority, CNCAN, about the intention to build the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) at the Mioveni nuclear platform, the Fostering ALFRED Construction (FALCON) international consortium – leading the project – and CNCAN agreed to initiate a formal pre-licensing phase for the project. This choice was deemed...
Using regression analysis of the LCOE values published by vendors of SMR projects and own calculations, effects of the scale, “learning” and reactor technology on the economics of SMR were identified. The competiveness of SMR projects versus another power sources was considered in relation to two markets - "Off-Grid" market (the power supply of remote communities) and "Grid" market (basic...
The Fast Breeder Test Reactor (FBTR) is sodium cooled, loop type fast reactor commissioned in the year 1985. It is a research reactor with an evolving core with a unique Pu/U mixed carbide fuel. The reactor has continued to provide valuable operating experience to improve its performance figure year after year. FBTR has been operated at different power levels up to 32 MWt/8 MWe over the last...
Within the GIF sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aim of the SO project includes (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification...
The paper explores options for converting Russian nuclear energy system (NES) to a two-component NES with a centralized closed nuclear fuel cycle (NFC), taking into account the uncertainty of the future. In order to take into account various trends in the development of NES, three groups of development scenarios were identified. The first group is growing scenarios in which the number of units...
Fast Breeder Test Reactor (FBTR), the fore-runner to the second stage of Indian nuclear power programme, is a 40 MWt / 13.2 MWe sodium cooled research reactor. FBTR first reached its criticality in October 1985 with a small core of 22 fuel subassemblies. The reactor uses indigenously developed Plutonium-Uranium mono carbide as the driver fuel and the fuel has seen a burn-up of 165 GWd/t...
The level of VVER and BN technology readiness in Russia provides the possibility to implement a two-component nuclear energy system based on new-generation VVER and BN -type reactors with a closed nuclear fuel cycle to solve problems of reducing the consumption of natural resources and spent nuclear fuel handling.
The paper presents the results of scenario-dynamic analysis of two-component...
A specific feature of subcritical once-through sodium-water steam generators operating as part of the BN-600 reactor power unit is the process of precipitation of nonvolatile impurities of the third circuit coolant in the steam generator evaporator, thus, forming deposits on the heat transfer surfaces. These impurities consist of corrosion products, various non-volatile salts and bases,...
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...
The Prototype Fast Breeder Reactor (PFBR) Steam Generator (SG) is periodically inspected using the indigenously developed robotic device PFBR SG inspection system. During inspection, if any tube is found to have an unacceptable degradation, then the tube has to be isolated from service. In the design of PFBR Steam Generator, a 10% margin is provided in the heat transfer area and thus SG tube...
Fast reactors (hereafter, FR) are a unique technological solution, which permits to provide enhancement of efficiency of uranium resource efficiency within the framework of closed fuel cycle. The FR is attractive less as a power reactor than as a basis to close the nuclear fuel cycle for potential customers.
The Treaty on the Non-Proliferation of Nuclear Weapons, International Atomic Energy...
Fast test reactor BOR-60 is one of Russia’s and world’s leading experimental facilities for testing a wide range of fuels, absorbing and structural materials intended for promising reactors.
In December 2019, BOR-60 celebrated its 50-year anniversary.
Nowadays and for the near future, BOR-60 is the only fast test reactor with unique experimental capabilities to perform research in various...
One of the conditions for the large-scale development of nuclear power (NP) in the world energy mix is the solution of the problem of non-proliferation of nuclear fissile materials.
To solve this problem, both institutional measures (e.g. the country's membership in the IAEA, IAEA inspection activities) and technological support measures (e.g. restrictions on the use of highly enriched...
At the IPPE JSC, the BR-10 research reactor was completely shut down in December 2002 after 43 years of operation. In 2003, the last reactor core was unloaded. At present, BR-10 is operated in the final shutdown mode in order to process accumulated spent radioactive waste of alkaline coolants.
Over the years of operation, BR-10 accumulated approximately 18-19 m3 of alkali metals (Na of the...
INPRO has made a brief historical overview of the IAEA activity in double- or multi-component nuclear power systems. Such systems include thermal neutron reactors and fast neutron reactors (FR) with a closed fuel cycle. Until the beginning of the 21st century the IAEA served mainly as a platform for information exchange in reactor development and deployment with focus on technical aspects of...
The application of Generation IV reactors offers improved sustainability of nuclear energy production and extends the current reserves while it helps to reduce the amount of nuclear waste. Therefore multiple demonstrator reactor designs are now under development. In fast reactors, due to its characteristic spectra and the reactor design, the effects of the leakage is much higher, which suggest...
Sodium-cooled Fast Reactor (SFR) possesses highest technology readiness level for deployment among six Gen-IV nuclear reactor designs intended to provide a low-carbon energy option and endure higher operating temperatures for longer service life (60-80 years). Thus, advanced materials developed for Gen-IV reactors should be able to withstand the harsh operating conditions allowing for safety...
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...
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,...
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, 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...
When transferring to two-component nuclear energy system at the initial stage of closing the nuclear fuel cycle of Russia, the number of fast breeder reactors BN type fuelled with plutonium do exist to a small extent and stocks of separated plutonium already are sizable and continue to grow due to pilot reprocessing SNF of VVER. Therefore, for this period of time, it is proposed to abandon...
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...
Plasma nitriding is considered as a plausible alternate hard-facing technology for enhancing the wear and fatigue resistance of large size and intricate fast reactor components. Plasma nitriding is an environmentally clean process that can be adopted in principle, to produce high surface hardness in a controlled manner, with little distortion of finished components. In this regard, an attempt...
Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commision), a methodology is proposed to enable the early consideration of safety in the undergoing design process. Before the use of mechanistic tools (CATHARE, SIMMER, APOLLO, etc.) that usually requires an advanced knowledge of the reactor design, this...
The results of Phase-1 neutronics benchmark analysis of CEFR start-up tests was presented in the 2nd Research Coordination Meeting (RCM-2), held at Beijing, China during 28th October to 1st November 2019. Participation of this benchmark exercise has provided us a wide international forum to inter-compare the method, computer codes and cross section data employed at IGCAR for the physics design...
Support ring is a key heavy structural component of the sodium-cooled Fast Reactor (SFR), which supports the whole reactor vessel and the reactor internals, are subjected to high pressure and high temperature and other wind loads, seismic loads, dead loads etc. Therefore, the security and stability of the support ring is essential to nuclear reactor. As the support ring has a super large...
The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-Pu-MA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and...
To improve the safety of future sodium cooled fast reactor, alternative Decay Heat Removal (DHR) path is envisaged through secondary sodium main circuit. Secondary Sodium Decay Heat Removal System (SSDHRS) transfers heat from secondary sodium to ambient through Air Heat Exchanger (AHX). SSDHRS is planned to operate in forced circulation mode using Class 3 power supply. SSDHRS in addition to...
AZNHEX is a deterministic code that solves the neutron diffusion equation with Hexagonal-Z geometry. It is part of the AZTLAN Platform, an initiative gathering the leading Mexican universities and nuclear energy research centers aimed to position Mexico in the midterm as a nuclear code developer for reactor core analysis and design. To tackle the challenges of a small core, like CEFR, with...
Adopting the 60-year design is regarded as one of the most effective means for the practical realization of Sodium-cooled Fast Reactor (SFR), which improves the economic efficiency and reduces the radioactive waste of SFR. In addition, since the happening of the severe accident (SA) at the Fukushima Daiichi Nuclear Power Station, the structural integrity evaluation of SA has been emphasized on...
The OECD/NEA Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) was launched in 2015 to study reactivity feedback coefficients and their uncertainties for a medium-sized 1,000 MWth metallic core and a large 3,600 MWth oxide core. In addition to investigations of the full core level, stand-alone multi-scale...
China already has 48 nuclear power reactors in operation, and 12 nuclear power reactors under construction till June，2020. The accumulated spent fuel in China has exceeded 8,500 tons. China implements the established policy of closed nuclear fuel cycle for the sustainable development of nuclear power. However, there seems no feasible development plan and road map to initiate and deploy a...
The Fukushima accident highlighted the need to improve the design of the safety systems in order to cope the consequences of a full blackout accident. In particular, the decay heat removal function must always be ensured in order to guarantee the integrity of the first barrier and the second one confining the radioactivity. The paper shows the main features of the pre-design of an innovative...
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...
The current stage of nuclear power development is often characterized by directly opposite trends: on the one hand, there is a growing understanding of the need to transition to a closed fuel cycle, and on the other hand, there is a very large uncertainty about the growth rates of nuclear power capacity. In such conditions, the main driver of the transition to SNF is not the need for capacity...
During the last decades, the development of new powerful computers and high performance analytical tools, along with the reduction of the approximations due to new methods implemented in the algorithms for the solution of the transport equation, pushed nuclear cross-sections data as the main source of uncertainty in neutronic calculations. This points out the importance in quantifying nuclear...
316H austenitic stainless steel is widely used in the manufacture of nuclear reactor components owing to its excellent comprehensive properties, such as main vessel, support assembly, etc. During the fusion welding of austenitic stainless steel, the tendency of hot cracking tends to occur when the structural restraint is too large. For preventing the cracking, it is usually desirable to form a...
Lead cooled reactor BREST-OD-300 design is under development as a part of Russian federal project "PRORYV". Two circuits are used for heat removal from the reactor. The special feature of two-circuit heat removal system is the potential risk of steam ingress into the core in case of large leak in the steam generator as a result, for example, of steam generator tube rupture (SGTR). The main...
Within the Generation IV International Forum, the partners of the Sodium-cooled Fast Reactor System Arrangement (China, Euratom, France, Japan, Korea, the Russian Federation, The United Kingdom and the United States of America) completed an evaluation of SFR advanced fuel options. This work was based on a preliminary work performed in the GIF SFR Advanced Fuel Project. It entailed a comparison...
ESFR-SMART belongs to the family of Gen-IV sodium cooled reactors. For its safety performance demonstration the detailed knowledge of selected safety related parameters is required. It is very important to know not only their spatial distribution and amplitude but also their mutual interdependence. In this paper the sodium void effect, the Doppler effect and fuel and cladding density effects...
The future of nuclear power lies in inherently safe fast reactors operating in a closed nuclear fuel cycle. The concept of inherent safety, which is based on deterministic exclusion of the most serious accidents due to the internal properties of the reactor, and not by creation of engineering barriers, is the foundation for ensuring the safety and economic efficiency of future nuclear power....
Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor, operating at Kalpakkam. FBTR core is originally designed to operate at 40 MWt using mixed oxide (MOX) fuel with 30% PuO2 and 70% UO2 (85% enriched U). However, due to the non-availability of enriched uranium, mixed carbide fuel (Mark-I (70% PuC+30%UC)) was chosen for the initial core. After first criticality in 1985,...
The influence of liquid lead on mechanical properties of ferritic-martensitic steel T91 and austenitic stainless steel 316L have been studied in the JRC’s LIquid Lead LAboratory (LILLA). LILLA allows testing of mechanical and corrosion properties of materials in liquid lead with controlled dissolved oxygen concentrations and for temperatures up to 650°C. The load is generated by pneumatic...
The MSFR (Molten Salt Fast Reactor) consists in a concept of high power molten salt reactor. In molten salt reactors, the fissile and fertile nuclei are dissolved in a circulating salt that acts as fuel and coolant. The physical state of the fuel permits to consider draining as a way to mitigate hypothetical accidents. Contrary to solid fuel FNR (Fast Neutron Reactor) concepts, in the MSFR,...
Within the framework of the new evolutive French 400MWth Advanced Sodium Technological Reactor for Industrial Demonstration project (ASTRID), qualification tests have been performed in order to demonstrate and confirm the feasibility and the performance of major components.
One of them is related to the Intermediate Heat Exchanger (IHX) and specifically about the interface between the inner...
The purpose of this work was to study and identify the features of the formation of external radiation doses to personnel during the production of a new mixed uranium-plutonium nitride fuel, which surpasses MOX fuel in almost all thermal and physical characteristics, the production of which is currently being developed in Russia.
Quite a few works have been devoted to the radiation...
To date, French nuclear power plants mainly use uranium extracted from the mines and enriched. The used fuel is reprocessed to extract useful materials, such as plutonium, which is recycled once in dedicated PWRs of the park. Once used, the quality of the plutonium decreases and it becomes more difficult to recycle this plutonium in thermal spectrum reactors.
A molten salt reactor (MSR)...
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...
For design and safety justification of fast reactors with liquid metal coolants, a complex of more than 20 experimental facilities of various functions and purposes, equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological facilities, has been created at IPPE JSC.
In addition IPPE JSC has a complex of fast neutron facilities, including two...
Fast Flux Test Facility (FFTF) was a research Sodium-cooled Fast Reactor (SFR) operated in the 1980’s with a goal to demonstrate the inherent and passive safety characteristics of the SFR design. In the frame of this study, an attempt has been made to develop a model of the FFTF suitable to predict the system’s behavior during the Unprotected Loss of Flow (ULOF) accident. In particular,...
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...
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...
Severe accident source term problem is the calculation of transport and release of radionuclides (RNs) from the reactor core to the environment for a hypothetical accident sequence of low probability. Generally, the source term problem is decoupled into four source terms: 1) In-vessel source term, 2) Interface source term 3) In-containment source term and 4) Environmental source term.
In...
In order to develop numerical modeling of hydrodynamics and heat exchange for substantiation of design and safety characteristics of fast reactors planning and carrying out experiments, generalizing the experimental results according to the physical laws and to their closing relations and verification of codes are required. To solve these problems, a complex of liquid metal facilities,...
Severe accidents in nuclear power industry (Three Mile Island, Chernobyl) brought NIKIET to the development of a fast lead-cooled reactor (BREST) concept at the end of the 1980s, which dramatically changed approaches to safety. To overcome contradictions between safety requirements and economical efficiency when designing a new-generation fast reactor, a new approach to selecting fundamental...
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...
In recent years, small power modular reactors (SMRs) have begun to be developed in many countries around the world. At the last IAEA meeting of Technical Working Group on Fast SMRs (September, 2019), several concepts for SMRs with different coolants were presented. Interest in such reactors is due to a number of their inherent features: reducing the construction term, a higher level of safety,...
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...
Three Work Packages (WPs) were defined in this Coordinated Research Project (CRP) whose objective was to estimate fission-product-transportation behaviour inside the reference pool-type sodium-cooled fast reactor (SFR) volumes (i.e., in-primary vessel, cover gas system and in-containment building) at different time scales under severe accident conditions. This WP, WP-2, is defined to estimate...
A preliminary core design of 100MWe Advanced Nitride-fueled Simplified (ANTS) Lead-bismuth cooled Fast Reactor (LFR) for civilian multi-cycle nuclear power plant has been investigated. The prime design constraint is on the core size with the active core diameter and height equal to 2.4 m and 1.0 m, respectively. The core is composed of 144 hexagonal fuel assemblies enclosed with 15-15Ti steel...
The main objective of the coordinated research project (CRP) on is to simulate the fission product transportation behavior of the reference pool-type sodium-cooled fast reactor (SFR) of 1250 MWth capacity with mixed oxide fuel under severe accident conditions. The accident considered is an unprotected loss of flow accident resulting in a core damage event with release of radionuclides. 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...
The purpose of this paper is to present the Mechanisms Engineering Test Loop (METL) facility installed at U.S. DOE’s Argonne National Laboratory located outside of Chicago, Illinois. The METL facility is a multipurpose intermediate-scale 750-gallon (2839l) sodium test facility used for the testing of liquid metal systems and components in prototypic reactor-grade sodium conditions. Some...
The accident at the Fukushima NPP (Japan) showed that the design of the power unit should consider the unforeseen excess of the external influence intensity. The accidents for internal reasons can be predicted on the basis of knowledge and it depends, first of all, on the designers approach, but it is impossible to foresee catastrophic external influences.
This method is proposed the heat...
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 Fast Breeder Test Reactor (FBTR) at Kalpakkam is the flagship of the second stage of the three-stage nuclear program of Department of Atomic Energy in India. Health physics services commenced at FBTR in April 1985 itself and the reactor went critical in October 1985. FBTR is a unique reactor utilizing U-Pu C as the fuel. Presently, FBTR has been operated upto a maximum power of 32 MW(t)...
The concept of a low-power 25 MW thermal reactor with a hard neutron spectrum in the core is proposed. The reactor has the following feature: a sufficiently high neutron flux density 3,4•1015 n/(sм2•s) in the center of the core, high average neutron energy 0,86 МeV in the center of the core, as well as a high proportion of hard En>0,8 МeV, neutrons 35%. Extremely high design parameters of the...
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...
This paper present the core optimization of the 1000 MW (electric) French commercial Sodium Fast Reactor, aimed at selecting few optimal configurations with respect to both core safety and reactor cost criteria. The Generation-IV reactor core design process must conciliate multiple goals (e.g. highest reactor safety levels in any situation, easy exploitability, affordable cost), which are...
Advanced nuclear technologies will play a significant role in meeting the growing demand for clean energy. To support the deployment of these technologies and enable long term innovation, the U.S. DOE has initiated a project to build a Versatile Test Reactor (VTR) with a compelling and urgent mission: testing innovative fuels, materials, sensors and instrumentation for various advanced...
As part of achieving sustainable development, a concept of the hybrid micro modular reactor (H-MMR) has been proposed by integrating the MMR design developed by KAIST with renewable energy and energy storage systems (ESS). The reactor power is designed to be 18MWth, and it is aimed for an ultra-long core lifetime for more than 20 years without refueling. The H-MMR core consists of 18 hexagonal...
Development and justification of innovative nuclear power technologies, spent nuclear fuel management and closed nuclear fuel cycle are the main issues that nuclear power industry faces nowadays.
Solving those tasks will be possible only with the help of modern research infrastructure permitting to conduct a full-scale experiments and to model various operating conditions. Russia is...
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...
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 that will support the development of advanced reactor technologies. This reactor will accelerate the irradiation testing of advanced nuclear fuels, materials, and potentially other components. The development efforts are structured in several phases to...
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...
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...
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...
