scholarly journals Effect of Fuel Burnup History on Neutronic Characteristics of WWER-1000 Core

2014 ◽  
pp. 14-18
Author(s):  
Yu. Ovdiienko ◽  
M. Yeremenko ◽  
V. Khalimonchuk ◽  
A. Kuchin ◽  
Yu. Bilodid
Keyword(s):  
Cross Sections ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Cycle Duration ◽  
Fuel Loading ◽  
Reactivity Coefficients ◽  
History Effect ◽  
Neutronic Characteristics ◽  
Nuclide Composition

In preparation of few-group cross-section libraries to be used in WWER macro-calculations, change in the nuclide composition during fuel burnup is commonly defined under invariable characteristics, averaged over the entire core (power, fuel and moderator temperature, moderator density etc.). In reality, conditions of fuel burnup are changing and this factor affects the fuel nuclide composition (the so-called spectral history effect). To account for real burnup history, it is necessary to take into consideration the dependence of cross-sections not only on burnup but also on history of neutron spectrum in the fuel burnup process. This paper analyzes fuel burnup history effect on neutronic characteristics of WWER-1000 core with use of the DYN3D code. The DYN3D code employs the local Pu-239 concentration as an indicator of burnup spectral history. The calculations have been performed for the first four fuel loadings of Khmelnitsky NPP unit 2 and stationary fuel loading with TVSA. The effect of fuel burnup history is shown both on macro-characteristics of the reactor core (boric acid concentration, fuel cycle duration, reactivity coefficients) and on local values of burnup and power.

Possibility of Different Fuel Cycles Usage in GT-MHR

2002 ◽  
Author(s):  
N. Kodochigov ◽  
Yu. Sukharev ◽  
E. Marova ◽  
N. Ponomarev-Stepnoy ◽  
E. Glushkov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Block Design ◽  
Average Power ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Fuel Load ◽  
Fuel Cycles ◽  
Partial Cycle ◽  
Control Rods ◽  
Neutronic Characteristics

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

scholarly journals Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

2019 ◽  
Vol 5 (4) ◽  
pp. 289-295 ◽  
Author(s):  
Olga I. Bulakh ◽  
Oleg K. Kostylev ◽  
Vladimir N. Nesterov ◽  
Eldar K. Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Alternative Energy ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Calculated Data ◽  
Fuel Burnup ◽  
The Core ◽  
High Temperature Reactors ◽  
High Temperature Gas

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

scholarly journals Analysis of the influence of nuclear fuel burnup on the 16N formation rate in the primary coolant of the WWER-1000 reactor

2021 ◽  
Vol 22 (1) ◽  
pp. 48-55
Author(s):  
Yu. Fylonych ◽  
◽  
V. Zaporozhan ◽  
O. Balashevskyi ◽  
K. Merkotan ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Critical Concentration ◽  
Formation Rate ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Primary Coolant ◽  
The Core ◽  
Fuel Assemblies ◽  
Water Volumes

The developed model of the WWER-1000 reactor using MCNP6.2 (Monte Carlo N-Particle Transport Code) includes the detailed core taking into account the design of the fuel assemblies, as well as the baffle, the lower plenum, the fuel support columns, the core barrel, a downcomer, and the reactor pressure vessel. It allows implementing multifunctional calculations such as recriticality with various fuel configurations, the critical concentration of boric acid, determination of the axial and radial peaking factor in the reactor core, etc. For obtaining the more precise result of the cumulation nitrogen-16 formation rate, the contribution from different water volumes was taken into account: in the core, above the fuel and the top nozzle, in the top nozzle of the fuel assembly, in the bottom nozzle, between the fuel and the bottom nozzle, in the axial channels of the baffle, in the reflector. In order to obtain the realistic boundary conditions, the change of the isotopic composition in the fuel assemblies during one fuel cycle was calculated using the ORIGEN-ARP of SCALE software. Therefore, the influence of the nuclear fuel depletion of fuel assemblies in the WWER-1000 reactor on the change of the basic neutron-physical characteristics was determined such as the distribution of the neutron flux density with the energies necessary to initiate the 16O(n,p)16N reaction, the average number of neutrons per fission, the neutron spectrum and average fission energy. As a result, the dependence of the nitrogen-16 formation rate in the primary coolant system on the nuclear fuel burnup is obtained.

Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

1981 ◽  
Vol 103 (4) ◽  
pp. 627-636 ◽  
Author(s):  
B. M. Ma
Keyword(s):  
Reactor Core ◽  
Fatigue Cracking ◽  
Computer Code ◽  
Fuel Burnup ◽  
Fuel Pellet ◽  
Gas Release ◽  
Unsteady State ◽  
Fission Gas ◽  
Fuel Elements ◽  
Fission Gas Release

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

scholarly journals Uncertainty Analysis of WWER-1000 Core Macroscopic Cross Sections due to Spectral Effects

2020 ◽  
pp. 39-46
Author(s):  
О. Kukhotska ◽  
I. Ovdiienko ◽  
M. Ieremenko
Keyword(s):  
Energy Spectrum ◽  
Isotopic Composition ◽  
Uncertainty Analysis ◽  
Neutron Flux ◽  
Flux Density ◽  
Cross Section ◽  
Cross Sections ◽  
Fuel Burnup ◽  
Thermophysical Parameters ◽  
Neutron Flux Density

The paper presents the results of uncertainty analysis of WWER‑1000 core macroscopic cross sections due to spectral effects during WWER‑1000 fuel burnup and the analysis of cross section sensitivity from thermophysical parameters of the calculated cell, which affect energy spectrum of neutron flux density. The calculation of changes in the isotopic composition during burnup and the preparation of macroscopic cross sections used the developed HELIOS computer model [1] for TVSA, which is currently operated at most Ukrainian WWER‑1000 units. The GRS approach applying Software for Uncertainty and Sensitivity Analyses (SUSA) [2] was chosen to assess the uncertainty of the macroscopic cross sections due to spectral effects and analysis of cross section sensitivity from thermophysical parameters. The spectral effect on macroscopic cross sections was taken into account by calculating the fuel burnup for variational sets of thermophysical parameters (fuel temperature, coolant temperature and density, boric acid concentration) prepared in advance by the SUSA program, as a result of which fuel isotopic composition vectors were obtained. After that, neutronic constants for the reference state were developed for each of the sets of isotopic composition, which corresponded to a certain set of thermophysical parameters. At the next stage, the uncertainty of macroscopic cross sections of the interaction due to the spectral effects on the isotopic composition of the fuel was analyzed using SUSA 4, followed by the analysis of cross section sensitivity from thermophysical parameters of the calculated cell affecting energy spectrum of neutron flux density. In the future, the uncertainty of two-group macroscopic diffusion constants can be used to estimate the overall uncertainty of neutronic characteristics in large-grid core calculations, in particular, in the safety analysis.

scholarly journals Investigation of Coolant Local Hydrodynamics in the Mixed Core of the VVER Reactor

2020 ◽  
Vol 63 (2) ◽  
pp. 151-162
Author(s):  
S. M. Dmitriev ◽  
A. V. Gerasimov ◽  
A. A. Dobrov ◽  
D. V. Doronkov ◽  
A. N. Pronin ◽  
...  
Keyword(s):  
Cross Sections ◽  
Experimental Studies ◽  
Reactor Core ◽  
Cross Flow ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Computational Hydrodynamics ◽  
Cross Section Area ◽  
Fuel Assemblies ◽  
Flow Pressure

The article presents the results of experimental studies of the local hydrodynamics of the coolant flow in the mixed core of the VVER reactor, consisting of the TVSA-T and TVSA-T mod.2 fuel assemblies. Modeling of the flow of the coolant flow in the fuel rod bundle was carried out on an aerodynamic test stand. The research was carried out on a model of a fragment of a mixed core of a VVER reactor consisting of one TVSA-T segment and two segments of the TVSA-T.mod2. The flow pressure fields were measured with a five-channel pneumometric probe. The flow pressure field was converted to the direction and value of the coolant velocity vector according to the dependencies obtained during calibration. To obtain a detailed data of the flow, a characteristic cross-section area of the model was selected, including the space cross flow between fuel assemblies and four rows of fuel rods of each of the TVSA fuel assemblies. In the framework of this study the analysis of the spatial distribution of the projections of the velocity of the coolant flow was fulfilled that has made it possible to pinpoint regularities that are intrinsic to the coolant flowing around spacing, mixing and combined spacing grates of the TVSA. Also, the values of the transverse flow of the coolant caused by the flow along hydraulically nonidentical grates were determined and their localization in the longitudinal and cross sections of the experimental model was revealed. Besides, the effect of accumulation of hydrodynamic flow disturbances in the longitudinal and cross sections of the model caused by the staggered arrangement of hydraulically non-identical grates was determined. The results of the study of the coolant cross flow between fuel assemblies interaction, i.e. between the adjacent TVSA-T and TVSA-T mod.2 fuel assemblies were adopted for practical use in the JSC of “Afrikantov OKB Mechanical Engineering” for assessing the heat engineering reliability of VVER reactor cores; also, they were included in the database for verification of computational hydrodynamics programs (CFD codes) and for detailed cell-based calculation of the reactor core.

A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations

2021 ◽  
Author(s):  
Jaakko Leppänen ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
Antti Rintala ◽  
Unna Lauranto
Keyword(s):  
Fuel Cycle ◽  
Natural Circulation ◽  
Reactor Core ◽  
District Heating ◽  
Research Centre ◽  
Computational Framework ◽  
Conceptual Design Phase ◽  
Burnable Absorber ◽  
Control Rods ◽  
Type Fuel

Abstract The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.

scholarly journals Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 581-590 ◽  
Author(s):  
Przemysław Stanisz ◽  
Jerzy Cetnar ◽  
Grażyna Domańska
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Nuclear Fuel Cycle ◽  
Minor Actinide ◽  
Closed Nuclear Fuel Cycle ◽  
European Collaboration ◽  
Multiple Recycling

Abstract The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

A Study of Transverse Buckling Effect on the Characteristics of Nuclides Burnup Wave in a Fast Neutron Multiplying Media

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Manish Kumar ◽  
Om Pal Singh
Keyword(s):  
Fast Neutron ◽  
Fuel Cycle ◽  
Fuel Burnup ◽  
Full Width ◽  
Half Maximum ◽  
Transient Time ◽  
Velocity Of Propagation ◽  
Neutron Leakage ◽  
Uranium Plutonium ◽  
Plutonium Fuel

A study of transverse buckling effect on the characteristics of nuclides burnup wave in multiplying media (cylindrical geometry) has been carried out. The burnup wave is characterized in terms of velocity of propagation, transient length (TL), and transient time (TT) in establishing the burnup wave and full width at half maximum (FWHM) in the established region of the wave. The uranium–plutonium fuel cycle is considered. The sensitivity of the results is studied for different radial buckling led leakage of neutrons. It is discovered that the velocity of the wave increases with the increase in the radius of the cylinder (i.e., reduction in the transverse buckling and hence increase in radial neutron leakage). FWHM is relatively insensitive to radial neutron leakage. The transient time and transient length are very large for smaller radius; these decrease with the increase in radius. The study provides insight on the build-up of burnup wave in the neutron multiplying media and brings out the importance of transverse buckling led radial neutron leakage on the characteristics of fuel burnup wave in multiplying media.

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