Effect of fuel particles' size and position variations on multiplication factor in pebble-bed nuclear reactors

Kerntechnik ◽  
2007 ◽  
Vol 72 (5-6) ◽  
pp. 251-254 ◽  
Author(s):  
L. Snoj ◽  
M. Ravnik
2014 ◽  
Vol 2014 (1) ◽  
pp. 17-22
Author(s):  
Abdelfettah Benchrif ◽  
◽  
Abdelouahed Chetaine ◽  
Hamid Amsil ◽  
◽  
...  

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá

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.


Author(s):  
Haitao Wang ◽  
Xin Wang

Spherical fuel elements with a diameter of 60mm are basic units of the nuclear fuel for the pebble-bed high temperature gas-cooled reactor (HTR). Each fuel element is treated as a graphite matrix containing around 10,000 randomly distributed fuel particles. The essential safety concept of the pebble-bed HTR is based on the objective that maximum temperature of the fuel particles does not exceed the design value. In this paper, a microstructure-based boundary element model is proposed for the large-scale thermal analysis of a spherical fuel element. This model presents detailed structural information of a large number of coated fuel particles dispersed in a spherical graphite matrix in order that temperature distributions at the level of fuel particles can be evaluated. The model is meshed with boundary elements in conjunction with the fast multipole method (FMM) in order that such large-scale computation is performed only in a personal desktop computer. Taking advantage of the fact that fuel particles are of the same shape, a similar sub-domain approach is used to establish the temperature translation mechanism between various layers of each fuel particle and to simplify the associated boundary element formulation. The numerical results demonstrate large-scale capacity of the proposed method for the multi-level temperature evaluation of the pebble-bed HTR fuel elements.


Author(s):  
Geoffrey J. Peter

The accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors considered for the next generation of Advanced High Temperature Reactors (AHTR), for nuclear power production, and for high-temperature hydrogen production using nuclear reactors to reduce the carbon footprint is examined in this paper. Blockages can cause local variations in flow and heat transfer that may lead to hot spots within the bed that could compromise reactor safety. Therefore, it is important to know the void fraction distribution and the interstitial velocity field in the packed bed. The blockage for this numerical study simulated a region with significantly lower void than that in the rest of the bed. Finite difference technique solved the simplified continuity, momentum, and energy equations. Any meaningful outcome of the solution depended largely upon the validity of the boundary conditions. Among them, the inlet and outlet velocity profiles required special attention. Thus, a close approximation to these profiles obtained from an experimental set-up established the boundary conditions. This paper presents the development of the elliptic-partial differential equation for a bed of pebbles, and the solution procedure. The paper also discusses velocity and temperature profiles obtained from both numerical and experimental setup, with and without effect of blockage. In addition, the paper compares the results obtained from the experimental set-up with numerical simulation using a commercially available code that uses finite element techniques.


Joule ◽  
2018 ◽  
Vol 2 (10) ◽  
pp. 1911-1914 ◽  
Author(s):  
Rainer Moormann ◽  
R. Scott Kemp ◽  
Ju Li
Keyword(s):  

2015 ◽  
Vol 1769 ◽  
Author(s):  
Rochkhudson B. de Faria ◽  
Felipe Torres ◽  
Fabiana B. A. Monteiro ◽  
Claubia Pereira

ABSTRACTSilicon carbide (SiC) has a potential to replacement zircaloy as fuel cladding material due to its high temperature tolerance, chemical stability and low neutron affinity. These characteristics may improve the economic and safety of nuclear reactors. Previous work has examined the possible use of SiC as a fuel cladding material in a PWR (Pressurized Water Reactor) environment. However, the advantage thermo mechanical and neutronic analysis replacement zircaloy cladding is not clear. Literature reviews has been done to predict the thermo mechanical behavior of SiC in high temperatures. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This codes system is widely accepted and used worldwide for safety analysis and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR. It was used the CSAS6 and TRITON modules. The goals are to evaluate the behavior of the infinite multiplication factor (kinf) and neutron flux using SiC as a fuel cladding material.


Author(s):  
I. V. Voitov ◽  
V. P. Kolos

The prospect of use of fuel in the form of micro particles (balls with a diameter about a millimeter formed by the fissile material and a protective cover to hold the radioactive fission products) in nuclear reactors is disclosed. It’s marked that flow ability, large specific surface of heat removal, extraordinary high resistance of micro fuel particles allow to design innovative safe reactors for various purpose: transportable, breeders, high-temperature, high neutron flux etc. It’s suggested to complete the active zone by bulk heat releasing assemblies. In them the advantages of spherical micro fuel particles and a coolant side supply to the bed through permeable distribution and branch channels are harmoniously combined in these assembles. It is presented the scheme of bulk assemblies and carried out the analysis of modeling of dynamics of a stream in permeable channels. It is shown that the mathematical description of liquid movement in such channels has ambiguity and discrepancy. To eliminate modeling shortcomings a new kinematic image of current in the permeable channels was offered. It was proposed instead of the existing one representing a jet to which particles of coolant were continuously joined or separated on the permeable wall. In the new interpretation the flow in the permeable channel is considered as turn of the stream at its simultaneous expansion or narrowing depending on there is outflow or inflow. On the base of this image the equation for determination of coolant pressure changing in the permeable channel is obtained; reaction of a stream for changing of flow rate increment is established, the tangent component of a velocity on a permeable wall is founded. Thereby the disadvantages of describing of coolant moving in the bulk assembles channels are eliminated. Permeable channels are used not only in nuclear reactors, but also in many other technological devices: catalytic reactors, heat exchangers, filters, collector and distributing systems. The obtained results can be used for designing other devices with permeable channels.


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