Passive safety systems are used in generation III+ evolutionary reactors and in generation IV advanced reactor designs, especially for the decay heat removal following an accidental event. These systems allow with one or more loops the heat transfer from the primary system to the external environment through the natural circulation of fluids or through boiling and condensation phenomena. A...
The technologies of information retrieval in a database with full-text semantic indexing are considered. The information retrieval process is considered as a cognitive-oriented process. The semantic image of the document context is presented as an ontology. An ontology is defined as a set of three interconnected systems (functional, conceptual and terminological), on which the operation of...
ะขhis work is carried out in order to assess the compliance of the radiation protection of personnel working at the complex experimental installations of JSC ยซSะกhEยป 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 ยฐC. This allows the ECFM to be deployed, for example, as part of the safety instrumentation in liquid...
Many elastomers seals are used in the nuclear industry. Among these elastomers, ethylene propylene diene monomer (EPDM) and silicone rubbers have excellent radiation stability. Both the rubbers can be used for gasket and O-ring application in Reprocessing Plants. To study the suitability of these rubbers for application in the plant, EPDM rubber compound and silicone rubber compounds were...
Here described a method for determination of mean prompt neutron lifetime in fast metal cores during critical experiments held in RFNC โ VNIIะขF using assembly machine FKBN-2. The evaluation of derivative using experimental dependence between asymptotic decrease coefficient and core parts gap ฮฑ(ะ) was proposed further to determination of the delayed critical state of the core. The value ...
The equipment supports are in constricted arrangement in the main vessel of the fast reactor. Under the condition of earthquake, equipment supports may sustain damage caused by the interaction between equipment supports and fluid, therefore, the evaluation of the fluid-structure interaction effect is an important aspect of the structural safety assessment of fast reactors. Using the added mass...
Characteristics of the molten-salt reactor-burner (MSR-burner) of minor actinides (MA), which are concentrated in spent nuclear fuel of power reactors, depend significantly on the physical-chemical properties of the fuel composition. In particular, the MA transmutation efficiency is mainly determined by the concentration of actinide fluorides in the molten-salt fuel composition [1]. In this...
In order to reduce the long-term potential hazard of waste from the reprocessing of spent nuclear fuel from thermal reactors and to increase the enviroment attractiveness of nuclear power in our country, work is underway to create a molten-salt reactor-burner of minor actinides (MSR-B). The first stage on this path is the creation of an investigative molten-salt reactor (IMSR) for testing key...
The report presents results of studies of the physicochemical characteristics of radioactive workplace aerosols formed during the production of mixed nitride uranium-plutonium fuel (activity particle-size distribution, nuclide composition, lung absorption type, elemental composition, reactive properties in the air). Taking into account these characteristics, the dose coefficients were...
The objective of this work is to reveal the dependences of the separation of americium-241 and uranium using sorption technology based on the solid-phase extractant TODGA. In technologies for the purification of radioactive waste of low and medium activity levels with low contents of actinides, sorption and ion exchange methods are widely used due to their high selective. The required...
The Versatile Test Reactor (VTR) is a reactor under development in the United States of America to provide a very high-flux fast neutron source. This reactor will accelerate the testing of advanced nuclear fuels, materials, and other potentially irradiated components. As this reactor design effort is underway to support eventual construction and operation, a necessary step is the development...
Characterization of alpha emitting nuclide and other fission products in the radioactive waste generated in reprocessing plants is a regulatory requirement for their disposal. The assay of plutonium in the solid radioactive wastes could be carried out either using gamma spectrometry or neutron counting, depending mainly on the surface dose of the container. Presence of large amount of fission...
In Russia, research is actively underway to develop a specialized molten-salt burner reactor (MSR-burner) of minor actinides (MA) from spent nuclear fuel of power reactors. Two candidate fluoride salts, LiF-BeF2 [1] and LiF-NaF-KF, are considered as the solvent of the reactor fuel components.
The purpose of the present paper is to study MA transmutation in the MSR-burner based on selected...
This paper aims at finding solutions of so important problems of nuclear power as decreasing the scope and the number of technological operations, as well as enhancing the proliferation resistance of fissile materials in nuclear fuel cycle by means of minimal changes in the cycle. The method is including fusion neutron sources with thorium blanket into future nuclear power system. In addition...
To determine uranium in the metallic phase in the presence of uranium oxide there is a reliable, so-called โbromine methodโ, which implies a metallic-oxide mixture treating in the bromine ethyl acetate solution. However, analogous manipulations with rare earth metals and their oxides do not provide such reliable data. Reduction melting of oxides in a graphite crucible with the melt composed of...
Programmable Logic Controllers offer complete automation solution and flexibility to control in a plant like the nuclear fuel reprocessing plant. To ensure the safety of both the plant and personnel, continuous monitoring and diagnostics of plant parameters are implemented through various means. Audible and visual alarms are provided to alert the operator in case of process abnormality. But...
Technology to utilize a dry recycled MOX powder has been developed as a part of MOX fabrication technology development for fast reactors. The purpose of this study is to develop a technology to control the density of MOX pellets with use of dry recycled MOX powder. A roll crusher and a jet mill were employed to prepare the recycled MOX powder which had three kinds of particle sizes (coarse,...
Melts based on the LiCl-KCl eutectic are becoming attractive in various industrial fields, including nuclear industries. However, their transport characteristics have not yet been sufficiently studied.
The purpose of this work is to study the electrical conductivity of melts similar to those formed during the dissolution of real nitride spent nuclear fuel in (LiCl-KCl)eut., and also to...
A pyrochemical technology for reprocessing spent nuclear fuel (SNF) and fast reactors is being implemented. One of the redistributions of pyrochemical technology is the electrochemical reduction of uranium dioxide (actinide oxides) with lithium in a LiCl - Li2O melt (1-2 wt.%) uranium dioxide and rare earth oxides at 650 ยฐC. To test the technological regimes of the reduction process, we used a...
In Purex process, Ruthenium is one of the troublesome fission products due to its complex chemistry and presence of multiple oxidation states &some extractable stable complexes in nitric acid medium. The higher concentrations of both stable and radioactive ruthenium isotopes, pose many challenges. During the reprocessing of FBR spent fuel, the tri and tetra nitrato complexes of ruthenium get...
New nuclear fuel cycles include reducing the long-term radiotoxicity of nuclear waste by separation and transmutation of long-lived transplutonium elements. Therefore, selective recovery of transuranic elements, especially actinides (III) โ americium and curium โ from high-level waste generated during spent nuclear fuel reprocessing is an important issue. Processes for extracting americium...
The present study demonstrates the use of inorganic ion-exchanger (IX) to condition the high-level liquid waste (HLW) by selective separation of one of the major radionuclide, cesium-137 (137Cs) from it. 137Cs possesses a broad range of potential applications in societal and agricultural area. In addition to this, the selective separation of 137Cs from HLW would drastically bring down...
In recent years, several countries, including Russia, have been developing a pyrochemical (anhydrous) method for spent nuclear fuel (SNF) reprocessing. Molten salts have several advantages, such as thermal and radiation stability, a wide electrochemical window, etc. They can be used practically at all technological stages of SNF processing.
The first stage of pyrochemical reprocessing...
Long-lived minor actinides (MA) like, Neptunium, Americium, and Curium are the major burden of nuclear power. Long-lived MAs have not yet been used as nuclear fuel. Therefore, the transmutation of long-lived MAs is introduced as an alternative to direct final disposal. In current work, we compare the performance of MA transmutation in a critical Single-fluid Double-zone Thorium-based Molten...
As a result of the operation of nuclear reactors, a certain amount of Cm is produced, which is included in the minor actinides series (MA). Among the long-lived Cm isotopes, Cm243 and Cm245 should be noted. Their fission cross section is over 2.5 barn. In this regard, Cm can be used as a fuel in a fast neutron nuclear reactor.
For the scientific research, was used a model of the RBEC reactor...
Due to the complexity of the fast reactor project and its technical uncertainty, the design needs a long period. In order to improve design and research ability of fast reactor and develop the technology of digital reactor, bring in the model-based systems engineering method for requirement analysis, function decomposition and architecture design and weigh the overall design, to the benefit of...
Molten Salt Reactors, as a whole reactor category, belong to the GenIV reactors. They can be designed as thermal, epithermal of fast systems for variety of applications. Especially the Molten Chloride Fast Reactors (MCFRs) provide very hard neutron spectra and very high neutron economy. Hence, MCFRs can be operated as breeders in the closed U-Pu and Th-U cycles or as breed-and-burn reactors in...
During the operation of the BN-800 reactor, a large amount of experimental data has been accumulated on critical states, the effectiveness of the control system, etc. It should be noted that the loading of the hybrid core in the initial period was constantly changing: different ratios of fuel assemblies with uranium fuel and MOX fuel, as well as the number of fuel assemblies in the...
The current energy production, resulting from the concepts related to the fission of fissile nuclides, nuclear energy, is of the order of 397,650 MWe produced by the 449 nuclear plants in operation and another 54,364 MWe to be supplied by another 54 under construction on the planet [2 ], data that demonstrate the growth in installed capacity and the installation of electrical energy from...
The paper presents the results of the end-to-end mathematical modeling of the BN reactor with integral equipment layout. The developed approach permits to validate RP characteristics and to study the process of the transfer of the predecessor of the delayed neutrons with the primary circuit coolant in the conditions of stratified current.
The approach includes a complex of specially developed...
The analysis and elaboration of the thermal-hydraulic characteristics based on the results the reactor plant (RP) commissioning allow validating the algorithms of passing the modes and sufficiency of the margins applied in the project related to the thermal-hydraulic characteristics of the main equipment.
The paper presents the comparative analysis of the start-up algorithms and operation...
ENEA has a long-lasting expertise in the design of Gen IV nuclear reactors, in particular the ones cooled by liquid Lead (LFRs). In the EU context, through the participation to the FALCON Consortium, ENEA is pursuing all the activities required to support the construction of ALFRED โ the European demonstrator of the LFRs โ in Romania.
S/U analyses are a paramount step for the licensing of...
A comprehensive CFD model of reactor pool of liquid sodium cooled pool type 600 MWe fast reactor design along with immersed reactor components is developed for detailed thermal hydraulic studies. Hot and cold pools along with immersed components represent the primary heat transport system. The two pools are physically separated by inner vessel, which completely envelopes the hot pool. Cold...
From a neutronic point of view, the effects of thermal expansion on the reactivity of a reactor core are an important feedback mechanism, both in steady-state and during many postulated accidents sequences. It is therefore necessary to model the expanded configuration in terms of shapes, densities and volumes as accurately as possible. Unfortunately, this is not easy for those regions that...
Fast reactor (FR) operation in closed nuclear fuel cycle (CNFC) is accompanied by the change in isotopic composition of a recycled fuel during a prolonged period of time (10-30 years). Series of similar calculations are required to determine optimal parameters of core charge and operational conditions during this transient phase of FR operation. To solve this problem, a RTM software complex...
The tally cards F6 and F7 in MCNP program allow users to calculate reactor power. After a time of operation, the fission products increased, which caused the delayed energy in the reactor. Thus, the power directly calculated by F6 and F7 would not correspond with the real value, and for the fast reactor, the energy distribution of fuel and other structural materials will also deviate from the...
For power production and 233U breeding from thorium, a preliminary neutronic design of an Accelerator-Driven Sub-critical System (ADS) is presented. The ADS reactor core design with โHEUโThorium Oxide fuelโ was coupled with proton accelerator and spallation target. The neutron source (ADS system) feasibility of HEU burning and isotopes production was evaluated. The multiplication factor Keff,...
The isotopic kinetics code BPSD is developed by IBRAE RAN in โCodes of New Generationโ subproject of โProryvโ project. BPSD solves fuel, absorber (boron carbide, dysprosium hafnate) transmutation, coolant (lead, sodium) and steel activation problems. Moreover, it carries out activation and residual heat calculations of materials. BPSD is intended to model materials, applied in fast reactors...
The computational code COMPLEX for radiation safety assessment of reactor and nuclear fuel cycle facilities is a set of programs (modules) combined by exchange data files and a pre- and post-processing system. The code is being developed in the "Codes of new generation" subproject of the "Proryv" project. The code includes the following modules:
- reactor core calculation modules based on...
High level liquid waste (HLLW) is often stored in large capacity horizontal cylindrical tanks especially in fast reactor fuel reprocessing plants. However, these huge tanks when partially filled, pose safety concerns due to seismicity. Violent sloshing during an earthquake-induced Fluid-Structure Interaction (FSI) can lead to catastrophic effects such as structural failures, gas entrainment...
The uncertainities of evaluated nuclear data represent one of the most important sources of uncertainity in the reactor physics simulation. The improvement of these data used is requiered for the development, safety assesment and licensing process of a reactor. Is generally recognised the need for further investigation (experimental included) regarding the uncertainities on some main...
This paper proposes a more stringent method for customizing project rules. This method customizes the comprehensive rules of the project and component reference database on the digital plant design platform based on some general design codes, standards and item classification principles in nuclear engineering, digitalization requirements in reactor design, plant layout, project management,...
Calculations of non-stationary processes of fast neutron reactors taking into account the spatiotemporal dependence of the neutron field is a rather complex process due to the significant influence on the calculation results of delayed neutrons, which make up a very small, less than a percent, part of all the neutrons of the reactor in its critical state. This circumstance is due to the fact...
The advanced development of NPE assumes gradual embedding fast neutron reactors, which enshures the most complete use of the uranium and thorium resources. Even on the theoretical level there are some alternative solutions for organization of operation of reactor facility and the fuel cycles. The experimental verification requirs the considerable amount of time and serious material resources....
To date spent nuclear fuel (SNF) reprocessing is a promising field of study. More than 370 thousand tons of SNF have been accumulated in the world and 10-12 thousand tons are added to this amount annually. In Russian Federation, ~25000 tons of SNF were accumulated according to the data of 2018. Pyrochemical technology of SNF processing, which is supposed to substitute aqueous technologies, is...
The National Institute for Nuclear Research (ININ) of Mexico, participates in the IAEA-CRP on Neutronics Benchmark of the Chinese Experimental Fast Reactor (CEFR) Start-Up Tests, which was proposed by China Institute of Atomic Energy (CIAE). The Mexican participation in this Benchmark is focused in two main goals: the first one, the use of SERPENT code for the generation of reference solutions...
The report presents the results of comparing the calculated data and readings of devices for monitoring water leakage into sodium, observed during a real leak in the BN-600 steam generator.
BN-600 implemented a section-modular scheme of a sodium-water steam generator. The damage of the heat exchange surface of the BN-600 steam generator occurred mainly in the initial period of plant operation...
CFD modeling was extensively used for the development of high-temperature furnaces for the carbothermal synthesis of uranium and plutonium nitrides and a furnace for the sintering of mixed nitride uranium-plutonium fuel pellets. This equipment is intended for use at the Pilot Demonstration Energy Complex (PDEC) being constructed in Seversk, Russia. The ะกFD-model of the carbothermal synthesis...
One of the most important problems in the further development of nuclear energy, from the point of view of its public acceptance, is the problem of safety. Thus, the development of new concepts for nuclear fission reactors with so-called "intrinsic safety" is a very urgent task. An equally important problem for the sustainable development of nuclear power is the need to expand the fuel base by...
