scholarly journals DESIGN-BASIS JUSTIFICATION FOR IMPLEMENTING TARGETED ENERGY RELEASE IN TEST OBJECTS OF THE IMPULSE GRAPHITE REACTOR

2020 ◽  
Vol 17 (2) ◽  
pp. 87-95
Author(s):  
V.A. Vityuk ◽  
Keyword(s):  
Fuel Assembly ◽  
Energy Release ◽  
Radial Distribution ◽  
Design Stage ◽  
Fuel Rods ◽  
Design Basis ◽  
Graphite Reactor ◽  
Fuel Assemblies ◽  
Technical Solutions ◽  
Irradiation Device

The article presents the approaches and technical solutions applied to assure preset axial and radial distribution of energy release in simulative fuel rods and fuel assemblies in the tests at the impulse graphite reactor. It considers the procedure for the design-basis justification of solutions that provide a given volume distribution of energy release in a test unit. The considerations are based on the example of heterogeneous fuel assemblies with the altitude separation of enriched fuel into two zones by a depleted uranium layer used to reproduce fissile material. The implementation of the procedure and making appropriate technical solutions made it possible to provide a targeted profile of the axial and radial distribution of energy release in a simulative fuel assembly at the design stage of an irradiation device. By the result of study, it is demonstrated that uniform radial energy release and targeted average energy release in upper and lower fission zones of experimental fuel assembly could be obtained at the level of 90.6 W/g (UO2) and 74 W/g (UO2), respectively. The measures include profiling of fuel pellets enrichment in fuel rods, using the pellets with an absorber at the zone ends, and a certain altitudinal positioning of the irradiation device in the reactor.

Implementation of Erbia With ZrB2 in 17×17 Assemblies for a 4-Loop Core With a 24-Month Cycle

2002 ◽  
Author(s):  
James E. Platte ◽  
Ernesto Pitruzzella ◽  
Youssef Shatilla ◽  
Baard Johansen
Keyword(s):  
Fuel Assembly ◽  
Temperature Coefficient ◽  
Cycle Length ◽  
Zirconium Diboride ◽  
Fuel Rods ◽  
Fuel Assemblies ◽  
Burnable Absorber

There are many types of burnable absorbers currently used in power reactors. They are used to provide reactivity and power peaking control. Westinghouse reactors most commonly use Zirconium Diboride Integral Fuel Burnable Absorbers (ZrB2) while Combustion Engineering reactors most commonly use Erbia Integral Fuel Burnable Absorbers (Erbia) in Combustion Engineering reactors. This paper documents the study to determine the effect of placing Erbia and ZrB2 within a Westinghouse 17×17 fuel assembly, and the effect of these ZrB2/Erbia assemblies on the physics characteristics of a representative Westinghouse 4-loop, 24 month cycle length design. The study consisted first of producing optimal within-assembly burnable absorber configurations where ∼25% of the ZrB2-bearing fuel rods within an assembly were replaced with Erbia-bearing fuel rods. This ratio was selected in order to provide an effective balance between potential peaking factor improvements and the known Erbia disadvantage of increased residual absorber penalty compared with ZrB2. The optimal patterns were selected as the ones that most reduced the assembly-wise cumulative peak-to-average rod power during the depletion compared with existing all-ZrB2 BA configurations with the same BA rod quantity loading. The second part of this study consisted of substituting various quantities of these ZrB2/Erbia feed fuel assemblies in a representative Westinghouse 4-loop, 24 month cycle core design to study the effect on power peaking factors, moderator temperature coefficient (MTC), and cycle length.

Advantages of Thorium Nuclear Fuel for Thermal-Neutron Reactors

2015 ◽  
Vol 1084 ◽  
pp. 275-279 ◽  
Author(s):  
Igor Shamanin ◽  
Sergey Bedenko ◽  
Ildar Gubaydulin
Keyword(s):  
Numerical Simulation ◽  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
Nuclear Fuel Cycle ◽  
Fuel Rods ◽  
Fuel Assemblies ◽  
Technical Solutions ◽  
New Generation ◽  
Thorium 232

The explanation of the effects which were found earlier in numerical simulation of parameters of open thorium-plutonium nuclear fuel cycle has been offered. Scientific and technical solutions allow considering the possibility of including thorium-232 in the fuel of nuclear reactors, which are based on existing design solutions, and beginning to design of new generation materials: a new generation of fuel rods and fuel assemblies, where the isotope uranium-238 will be completely replaced with thorium-232.

Materials of New Generation in Nuclear Power Industry

2014 ◽  
Vol 1040 ◽  
pp. 74-79
Author(s):  
Igor Shamanin ◽  
Sergey Bedenko ◽  
Ildar Gubaydulin ◽  
Nataliya Novikova
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
Nuclear Fuel Cycle ◽  
Fuel Rods ◽  
Fuel Assemblies ◽  
Technical Solutions ◽  
New Generation ◽  
Thorium 232

The results providing advantages of thorium-232 as a reproducing nuclide in comparison with uranium-238 as a part of nuclear fuel of new generation reactors are presented. The explanation of the effects which were found earlier in numerical simulation of parameters of open thorium - plutonium nuclear fuel cycle is offered. Scientific and technical solutions allow considering the possibility of including thorium-232 in the fuel of nuclear reactors, which are based on existing design solutions, and beginning the design of new generation materials: a new generation of fuel rods and fuel assemblies, where the isotope uranium-238 will be completely replaced with thorium-232.

scholarly journals Engineering Margin in Calculations of Energy Release in VVER-1000 core

2018 ◽  
pp. 20-26
Author(s):  
A.M. Abdullayev ◽  
A.I. Zhukov ◽  
S.V. Maryokhin ◽  
S.D. Riabchykov
Keyword(s):  
Energy Release ◽  
Three Dimensional ◽  
Measurement Data ◽  
Reactor Core ◽  
Small Scale ◽  
Two Dimensional ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies

A method for calculating the engineering margin factor (EMF) in calculations of the energy release in the core of VVER-1000 reactors is proposed in the paper. The analysis of various approaches in the calculation of EMF is carried out and various factors influencing EMF and the ways of their consideration —deterministic and statistical — are determined. The main attention is paid to the influence of gaps between the fuel assemblies on the energy release of fuel rods and the contribution of this factor to the EMF. The limitations and conservatism of two-dimensional small-scale calculations of the energy release of fuel rods in case of deviation of the gap size between the fuel assemblies from the design one are shown. A three-dimensional approach to calculating the contribution of gaps to the EMF is proposed. The approach is based on detailed measurements of the shape of fuel assemblies removed from the core performed at Zaporizhzhya NPP [13]; simulation of the distribution of gaps in the reactor core [16] using measurement data; two-dimensional calculations of the energy release of fuel rods in separate fuel assemblies, surrounded by gaps of different widths, with mirroring boundary conditions; three-dimensional calculations of energy release of fuel rods in fuel assemblies in the reactor core. Two-dimensional and three-dimensional calculations are performed by the wellknown ALPHA-H/PHOENIX-H/ANC-H codes. The proposed approach allows considering not only the change in the fuel rod power, particularly of the peripheral rods, which is inherent in the currently used methods of calculating EMF, but also takes into account the change in the power of the fuel assemblies in the core, which makes the proposed method more realistic and removes the excessive conservatism of EMF calculations and, thereby, allows improving fuel efficiency. For fuel assemblies produced by Westinghouse, it is proposed to use full EMF: for fuel rod power (FΔH) 1.111 and for fuel rod linear power (FQ) 1.173. The use of the BEACONTM monitoring system makes it possible to further reduce the EMF: for fuel rod power (FΔH) - up to 1.084 and for fuel rod linear power (FQ) - up to 1.121.

scholarly journals A METHODOLOGY TO SUPPORT COMPANIES IN THE FIRST STEPS TOWARDS DE-MANUFACTURING

2021 ◽  
Vol 1 ◽  
pp. 131-140
Author(s):  
Federica Cappelletti ◽  
Marta Rossi ◽  
Michele Germani ◽  
Mohammad Shadman Hanif
Keyword(s):  
Environmental Impact ◽  
End Of Life ◽  
Design Stage ◽  
Life Strategies ◽  
Disassembly Sequence ◽  
The Status ◽  
The Cost ◽  
Technical Solutions ◽  
First Time ◽  
Complex Activities

AbstractDe-manufacturing and re-manufacturing are fundamental technical solutions to efficiently recover value from post-use products. Disassembly in one of the most complex activities in de-manufacturing because i) the more manual it is the higher is its cost, ii) disassembly times are variable due to uncertainty of conditions of products reaching their EoL, and iii) because it is necessary to know which components to disassemble to balance the cost of disassembly. The paper proposes a methodology that finds ways of applications: it can be applied at the design stage to detect space for product design improvements, and it also represents a baseline from organizations approaching de-manufacturing for the first time. The methodology consists of four main steps, in which firstly targets components are identified, according to their environmental impact; secondly their disassembly sequence is qualitatively evaluated, and successively it is quantitatively determined via disassembly times, predicting also the status of the component at their End of Life. The aim of the methodology is reached at the fourth phase when alternative, eco-friendlier End of Life strategies are proposed, verified, and chosen.

scholarly journals Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident

Materials ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 1818
Author(s):  
Di-Si Wang ◽  
Bo Liu ◽  
Sheng Yang ◽  
Bin Xi ◽  
Long Gu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Steam Generator ◽  
Metal Flow ◽  
Maximum Temperature ◽  
Steam Generator Tube ◽  
Driven System ◽  
Fuel Assemblies ◽  
Accelerator Driven System ◽  
The Impact ◽  
Steam Bubble

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Application of CFD Model for Inlet Flow Region of 17×17 Fuel Assembly

2006 ◽  
Author(s):  
Milorad B. Dzodzo ◽  
Bin Liu ◽  
Pablo R. Rubiolo ◽  
Zeses E. Karoutas ◽  
Michael Y. Young
Keyword(s):  
Fuel Assembly ◽  
Fluid Dynamic ◽  
Flow Distribution ◽  
Flow Region ◽  
Fretting Wear ◽  
Jet Flows ◽  
Cfd Model ◽  
Lateral Velocity ◽  
Fuel Assemblies ◽  
Inlet Conditions

A numerical investigation was performed to study the variation in axial and lateral velocity profiles occurring downstream of the inlet nozzle of a typical Westinghouse 17×17 PWR fuel assembly. A Computational Fluid Dynamic (CFD) model was developed with commercial CFD software. The model comprised the lower region of the fuel assembly, including: the Debris Filter Bottom Nozzle (DFBN), P-grid, Bottom Inconel grid, one and half grid span, as well as the lower core plate hole. The purpose of the study was to obtain insight into the flow redistribution resulting from the interaction of the jet arising from the lower core plate hole and the fuel assembly structure. In particular the axial and lateral velocities before and after the nozzle were studied. The results, axial and lateral velocity contours, streamlines and maximum axial and lateral velocity distributions at various elevations are presented and discussed in relation to the potential risk of high turbulent excitation over the rod and the resulting rod-to-grid fretting-wear damage. The CFD model results indicated that the large jet flows from the lower core plate are effectively dissipated by DFBN nozzle and the grids components of the fuel assembly. The breakup of the large jets in the DFBN and the lower grids helps to reduce the steep velocity gradients and thus the rod vibration and fretting-wear risk in the lower part of the fuel assembly. The presented CFD model is one step towards developing advanced tools that can be used to confirm and evaluate the effect of complex PWR structures on flow distribution. In the future the presented model could be integrated in a larger CFD model involving several fuel assemblies for evaluating the lateral velocities generated due to the non-uniform inlet conditions into the various fuel assemblies.

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

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