scholarly journals Justification of VVER-1000 safety when using fuel compositions doped by protactinium and neptunium

2020 ◽  
Vol 6 (2) ◽  
pp. 99-104 ◽  
Author(s):  
Tuul Baatar ◽  
Evgeny G. Kulikov
Keyword(s):  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Accumulation Rate ◽  
Fission Products ◽  
Fuel Burnup ◽  
Fuel Composition ◽  
Gradual Improvement ◽  
Nuclear Power Development ◽  
Reactivity Coefficients ◽  
Water Reactors

Increasing fuel burnup is one of the important areas of nuclear power development. Currently, the most common type of light-water reactors is characterized by burnup ratios of about 5%, i.e., only a small fraction of fuel is used to generate electricity. The paper considers the possibility of a significant increase in fuel burnup due by introducing protactinium and neptunium into the fuel composition. The chains of nuclide transformations starting with protactinium and neptunium are characterized by a gradual improvement in the multiplying properties, which ensures increased fuel burnup. At the same time, a situation may be observed when the multiplying properties of a fuel composition are improved during the campaign, which indicates that at a certain point in time the accumulation rate of fissile nuclides from protactinium and neptunium exceeds the accumulation rate of fission products. While protactinium is hardly accessible in sufficient quantities, neptunium is contained in spent nuclear fuel, a significant amount of which is stored in on-site facilities. Therefore, from a practical perspective, the introduction of neptunium into fuel compositions seems to be more preferable. The novelty of the work is the analysis of the effects of protactinium and neptunium on the reactivity coefficients during fuel campaigns. The calculations were carried out for a VVER-1000 type reactor using the SCALE-6.2 software package.

scholarly journals Nuclear Fuel Cycle with Minimized Waste

2019 ◽  
pp. 30-35
Author(s):  
V. Moiseenko ◽  
S. Chernitskiy
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
Nuclear Fuel Cycle ◽  
Fission Neutron ◽  
Fuel Composition ◽  
Monte Carlo Code ◽  
Metallic Fuel

A uranium-based nuclear fuel and fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a ‘balanced’ fuel only uranium-238 content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The major fissionable component of the fuel is plutonium is chosen. This makes it possible to abandon the use of uranium-235, whose reserves are quickly exhausted. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. This feature simplifies maintaining the closed nuclear fuel cycle and provides its periodicity. In the fuel balance calculations, nine isotopes of uranium, neptunium, plutonium and americium are used. This number of elements is not complete, but is quite sufficient for calculations which are used for conceptual analysis. For more detailed consideration, this set may be substantially expanded. The variation of the fuel composition depending on the reactor size is not too big. The model accounts for fission, neutron capture and decays. Using MCNPX numerical Monte-Carlo code, the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components. In this way, it can be assumed that transuranic elements will constantly return to such a reactor, and only fission products will be sent to storage. This will significantly reduce the radioactivity of spent nuclear fuel. It is important that the storage time for the fission products is much less than for the spent nuclear fuel, just about 300 years.

scholarly journals On the possibility to improve mixed uranium-plutonium fuel in fast reactors

2020 ◽  
Vol 6 (2) ◽  
pp. 131-135
Author(s):  
Vladimir A. Eliseev ◽  
Dmitry A. Klinov ◽  
Noël Camarcat ◽  
David Lemasson ◽  
Clement Mériot ◽  
...  
Keyword(s):  
Isotopic Composition ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Light Water Reactors ◽  
Fast Reactors ◽  
Light Water ◽  
Fuel Cycles ◽  
Mox Fuel ◽  
Water Reactors ◽  
Operational Activities

Accumulation of plutonium extracted from the spent nuclear fuel (SNF) of light water reactors is one of the central problems in nuclear power. To reduce out-of-the-reactor Pu inventory, leading nuclear power countries (France, Japan) use plutonium in light water power reactors in the form of MOX fuel, with half of Pu fissioning in this fuel. The rest of Pu cannot be reused easily and efficiently in light water reactors because of the high content of even isotopes. Plutonium for which there are no potential consumers is accumulated. Unlike thermal reactors, fast reactors take plutonium of any isotopic composition. That makes it possible to improve plutonium isotopic composition and to reduce the fraction of even isotopes to the level that allows reuse of such plutonium in thermal reactors. The idea of changing the isotopic composition of Pu in fast reactors is well-known. The originality of the research lies in applying this idea to combine the fuel cycles of fast and thermal reactors. Pu isotopic composition can be improved by combining certain operational activities in order to supply fuel to thermal and fast reactors. Scientific and technological justification of the possibility will let Russian BN technologies and French MOX fuel technologies work in synergy with thermal reactors.

Distribution of Palladium During Spent Fuel Reprocessing

2003 ◽  
Author(s):  
Yu. Pokhitonov ◽  
V. Romanovski ◽  
P. Rance
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Platinum Group Metals ◽  
Spent Fuel ◽  
Platinum Metals ◽  
Spent Fuel Reprocessing ◽  
Platinum Group ◽  
Fuel Reprocessing

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile and fertile isotopes and facilitate disposal of waste elements. Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants (NPPs,) there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. Given current predictions for nuclear power generation, it is predicted that the quantities of palladium to be accumulated by the middle of this century will be comparable with those of the natural sources, and the quantities of rhodium in spent nuclear fuel may even exceed those in natural sources. These facts allow one to consider spent nuclear fuel generated by NPPs as a potential source for creation of a strategic stock of platinum group metals. Despite of a rather strong prediction of growth of palladium consumption, demand for “reactor” palladium in industry should not be expected because it contains a long-lived radioactive isotope 107Pd (half-life 6,5·105 years) and will thus be radioactive for a very considerable period, which, naturally, restricts its possible applications. It is presently difficult to predict all the areas for potential use of “reactor” palladium in the future, but one can envisage that the use of palladium in radwaste reprocessing technology (e.g. immobilization of iodine-129 and trans-plutonium elements) and in the hydrogen energy cycle may play a decisive role in developing the demand for this metal. Realization of platinum metals recovery operation before HLW vitrification will also have one further benefit, namely to simplify the vitrification process, because platinum group metals may in certain circumstances have adverse effects on the vitrification process. The paper will report data on platinum metals (PGM) distribution in spent fuel reprocessing products and the different alternatives of palladium separation flowsheets from HLW are presented. It is shown, that spent fuel dissolution conditions can affect the palladium distribution between solution and insoluble precipitates. The most important factors, which determine the composition and the yield of residues resulting from fuel dissolution, are the temperature and acid concentration. Apparently, a careful selection of fuel dissolution process parameters would make it possible to direct the main part of palladium to the 1st cycle raffinate together with the other fission products. In the authors’ opinion, the development of an efficient technology for palladium recovery requires the conception of a suitable flow-sheet and the choice of optimal regimes of “reactor” palladium recovery concurrently with the resolution of the problem of HLW partitioning when using the same facilities.

scholarly journals On a Long Term Strategy for the Success of Nuclear Power

2017 ◽  
Author(s):  
Bruno Merk ◽  
Dzianis Litskevich ◽  
Karl R. Whittle ◽  
Mark Bankhead ◽  
Richard Taylor ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Spent Nuclear Fuel ◽  
Nuclear Reactors ◽  
Theoretical Perspective ◽  
Strategic Development ◽  
Safety And Reliability ◽  
Water Reactors

The current generation of nuclear reactors are evolutionary in design, mostly based on the technology originally designed to power submarines, and dominated by Light Water Reactors. The aims of the GenIV consortium are driven by sustainability, safety and reliability, economics, and proliferation resistance. The aims are extended here to encompass the ultimate and universal vision for strategic development of energy production, the ‘perpetuum mobile’ – at least as close as possible. We propose to rethink nuclear reactor design with the mission to develop a system which uses no fresh resources and produces no fresh waste during operation as well as generates power safe and reliably in economic way. The results of the innovative simulations presented here demonstrate that, from a theoretical perspective, it is feasible to fulfil the mission through the reuse of spent nuclear fuel from currently operating reactors as the fuel for a new reactor. The produced waste is less burdensome than current spent nuclear fuel which is used as feed to the system. However, safety, reliability and operational economics will need to be demonstrated to create the basis for the long term success of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source.

scholarly journals Role of Passive Safety Systems in Chinese Nuclear Power Development

2009 ◽  
Vol 2009 ◽  
pp. 1-7 ◽  
Author(s):  
X. Cheng ◽  
Y. H. Yang ◽  
Y. Ouyang ◽  
H. X. Miao
Keyword(s):  
Nuclear Power ◽  
Passive Safety ◽  
Pressurized Water Reactors ◽  
Chinese Government ◽  
Power Development ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Nuclear Power Development ◽  
Water Reactors ◽  
Safety Systems

Passive safety systems have been widely applied to advanced water-cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear power development in China requires reactor concepts with high safety level. For the near-term and medium-term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. This paper describes some important criteria and the development program of the Chinese large-scale pressurized water reactors. An overview on representative research activities and results achieved so far on passive safety systems in various institutions is presented.

scholarly journals Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor

2019 ◽  
Vol 5 (4) ◽  
pp. 337-343
Author(s):  
Sergey N. Ivanov ◽  
Sergey I. Porollo ◽  
Yury D. Baranaev ◽  
Vladimir F. Timofeev ◽  
Yury V. Kharizomenov
Keyword(s):  
Dose Rate ◽  
Water Samples ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Fuel Composition ◽  
Spent Fuel ◽  
Artificial Defect ◽  
Corrosion Tests ◽  
Long Term Storage ◽  
Fuel Elements

Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings. We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments. Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage.

scholarly journals Assessment of the Radiotoxicity of Spent Nuclear Fuel from a Fleet of PWR Reactors

Energies ◽  
2021 ◽  
Vol 14 (11) ◽  
pp. 3094
Author(s):  
Mikołaj Oettingen
Keyword(s):  
Isotopic Composition ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Hybrid Energy ◽  
Cycle Simulation ◽  
Water Reactors

The paper presents the methodology for the estimation of the long-term actinides radiotoxicity and isotopic composition of spent nuclear fuel from a fleet of Pressurized Water Reactors (PWR). The methodology was developed using three independent numerical tools: the Spent Fuel Isotopic Composition database, the Nuclear Fuel Cycle Simulation System and the Monte Carlo Continuous Energy Burnup Code. The validation of spent fuel isotopic compositions obtained in the numerical modeling was performed using the available experimental data. A nuclear power embarking country benchmark was implemented for the verification and testing of the methodology. The obtained radiotoxicity reaches the reference levels at about 1.3 × 105 years, which is common for the PWR spent nuclear fuel. The presented methodology may be incorporated into a more versatile numerical tool for the modeling of hybrid energy systems.

scholarly journals Assessment of activity resources by nomenclature of main radiation-hazardous radionuclides during the long-term management of spent nuclear fuel of the Belarusian NPP

2020 ◽  
Vol 65 (2) ◽  
pp. 243-251
Author(s):  
J. A. Korchova ◽  
N. V. Harbachova ◽  
N. D. Kuzmina ◽  
N. V. Kulich
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Fuel ◽  
Management Strategy ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Fuel Management ◽  
Probabilistic Forecasting ◽  
The Government

The purpose of the study is calculation research of the radiation characteristics of fission products and actinides at different phases of spent nuclear fuel (SNF) management of the Belarusian Nuclear Power Plant (NPP). The study is aimed at the scientific support of the government decision as determined by the “On approval of the spent nuclear fuel management Strategy of the Belarusian nuclear power plant”. А probabilistic forecasting model and an effective code CUB for the spent nuclear fuel radioactivity inventory assessment were developed by the authors. Radionuclides activities as function of nuclear fuel burnup for nuclear fuel with the initial enrichment on the 235U equals to 4.81 % on the base of approximation relations of Regulation RB-093-14 (Moskow, 2014) have been calculated. Basic relations between specific activities of the main hazardous fission products and actinide, the SNF burnup and initial degree of fuel enrichment were analyzed. The rates of decrease of individual and total fission products and actinides activities of the Units №1 and 2 of the Belarusian NPP were obtained depending on the specific phase of spent SNF management. The results are of value for decision-making on ecology acceptable SNF management option introduced by Spent Nuclear Fuel Management Strategy of the Belarusian NPP.

scholarly journals THERMAL ANALYSIS OF DRY STORAGE AND TRANSPORTATION CASKS CASTOR USING COBRA-SFS

2020 ◽  
Vol 28 ◽  
pp. 32-41
Author(s):  
Martin Ševeček ◽  
Mojmír Valach ◽  
Chih-Hao Lee
Keyword(s):  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Construction Materials ◽  
Safety Evaluation ◽  
Fission Products ◽  
The Czech Republic ◽  
Dry Storage ◽  
State Office ◽  
The World ◽  
Residual Heat

State Office for Nuclear Safety is a regulator body in the Czech Republic handling also licensing of storage and transportation casks (SCs) of the spent nuclear fuel (SNF). One of the main concerns for the dry storage of SNF is the safe removal of the residual heat resulting from the decay of fission products, actinides, and activated construction materials. There are many storage systems in use around the world with different storage configuration, fuel designs and boundary conditions that need to be evaluated. Two thermal models of SNF storage casks CASTOR 440/84 and CASTOR 440/84M were developed using the COBRA-SFS code, which is an internationally accepted and validated code used for licensing of dry storage casks and safety evaluation. The casks are in operation at the Nuclear Power Plant Dukovany for dry storage of the WWER-440 SNF. Both models were validated and used for evaluation of several problems in different configuration and particularly for licensing purposes at the SUJB.

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