RESEARCH OF MODELS OF NEW PLATE HEAT-RELEASING ELEMENT

2020 ◽  
Vol 4 (1) ◽  
pp. 73-80
Author(s):  
Ilkhom Hikmatov ◽  
◽  
Fakhrulla Kungurov ◽  
Sapar Baitelesov ◽  
Davronbek Tojiboev ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Main Parameter ◽  
Research Reactor ◽  
Reactor Power ◽  
High Flow ◽  
Main Role ◽  
Research Reactors ◽  
Neutron Flux Density ◽  
Fuel Elements ◽  
High Neutron

The main parameter of research reactors is the neutron flux density. To obtain high neutron fluxes, the research reactor must be compact and the reactor power must be maximized. Nuclear fuel plays the main role in high-flow research reactors. Nuclear fuel using UO2is limited by the density of uranium in fuel elements (FUEL ELEMENTS) 3 g / sm3

Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor

2010 ◽  
Author(s):  
Aimin Zhang ◽  
Yalun Kang
Keyword(s):  
Fuel Element ◽  
Neutron Flux ◽  
Research Reactor ◽  
High Heat ◽  
Reactor Power ◽  
High Heat Flux ◽  
Water Tank ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Plate Type

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

Research Reactor VR-1 and Determination of Neutron Spatial Distribution in Its Radial Experimental Channel

2009 ◽  
Author(s):  
Marija Mileticˇ
Keyword(s):  
Spatial Distribution ◽  
Neutron Flux ◽  
Flux Density ◽  
Research Reactor ◽  
Reactor Power ◽  
Activation Detector ◽  
Experimental Channel ◽  
Neutron Flux Density ◽  
Radial Channel

The research reactor VR-1 is operated by Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University (CTU) in Prague. It is a pool-type, light-water reactor, with low enriched uranium. Maximum thermal power is 1kW (equal to 1·108 impulses/second when compared with reactors with higher power). Research on VR-1 reactor is mainly used for the education of university students, preparation and testing of new educational methodologies, investigation of reactor lattice parameters, reactor dynamics study, research in the control equipment field, neutron detector calibration, etc. One of the applications performed by students is the determination of the absolute value of the neutron flux density (also known as Neutron Spatial Distribution) in the radial experimental channel in reactor VR-1. The method used for this measurement is Neutron Activation Analysis. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector which is irradiated in the experimental channel. The activity of the produced radioactive products (radioisotopes) is then measured by means of appropriate counter system (in our case, High Purity Germanium detector). For this measurement totally 34 gold foils were irradiated at different reactor power levels and various positions in radial channel in aim to determine the neutron spatial distribution in radial channel. Interesting results about symmetry, value and dependence on reactor power level of neutron flux density were obtained.

Nondestructive evaluation of plate type nuclear fuel elements for PURNIMA-III and KAMINI research reactors

1992 ◽  
pp. 130-137
Author(s):  
J.K. Ghosh ◽  
S. Muralidhar ◽  
K.N. Chandrasekharan ◽  
V.D. Pandey ◽  
A. Subramanian ◽  
...  
Keyword(s):  
Nondestructive Evaluation ◽  
Nuclear Fuel ◽  
Research Reactors ◽  
Fuel Elements ◽  
Plate Type

scholarly journals Changes of the Neutron Flux of the Nuclear Reactor Triga Mark III Since the Conversion from High to Low 235U Enrichment

2021 ◽  
Vol 8 (2) ◽  
pp. 149-153
Author(s):  
C. Vázquez-López ◽  
O. Del Ángel-Gómez ◽  
R. Raya-Arredondo ◽  
S. S. Cruz-Galindo ◽  
J. I. Golzarri-Moreno ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Nuclear Research ◽  
Reactor Power ◽  
Uranium Fuel ◽  
Acrylic Sheet ◽  
Enriched Uranium ◽  
Reactor Operating ◽  
Fuel Elements

The neutron flux of the Triga Mark III research reactor was studied using nuclear track detectors. The facility of the National Institute for Nuclear Research (ININ), operates with a new core load of 85 LEU 30/20 (Low Enriched Uranium) fuel elements. The reactor provides a neutron flux around 2 × 1012 n cm-2s-1 at the irradiation channel. In this channel, CR-39 (allyl diglycol policarbonate) Landauer® detectors were exposed to neutrons; the detectors were covered with a 3 mm acrylic sheet for (n, p) reaction. Results show a linear response between the reactor power in the range 0.1 - 7 kW, and the average nuclear track density with data reproducibility and relatively low uncertainty (±5%). The method is a simple technique, fast and reliable procedure to monitor the research reactor operating power levels.

Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran Research Reactor

2016 ◽  
Vol 58 (9) ◽  
pp. 763-766 ◽  
Author(s):  
Mohammad Hosein Choopan Dastjerdi ◽  
Hossein Khalafi ◽  
Yaser Kasesaz ◽  
Amir Movafeghi
Keyword(s):  
Nuclear Fuel ◽  
Research Reactor ◽  
Neutron Radiography ◽  
Fuel Rods ◽  
Tehran Research Reactor ◽  
Nuclear Fuel Rods

USA–Iran, 2009–2010

2018 ◽  
Author(s):  
Nicholas J. Wheeler
Keyword(s):  
Nuclear Fuel ◽  
Research Reactor ◽  
Interpersonal Trust ◽  
Obama Administration ◽  
Face To Face ◽  
Tehran Research Reactor

This chapter examines the attempts by the first Obama Administration to reach out to Iran in an effort to build trust. It traces the failure of Obama’s diplomatic efforts to secure any reciprocation from Iranian leaders. The lack of reciprocation shows the problem of accurate signal interpretation when there is no trust. It focuses on the negotiations in 2009–10 over limiting Iran’s supply of nuclear fuel in return for refuelling the Tehran Research Reactor. The chapter argues these negotiations failed because of the lack of trust. What makes this case so important is that there was no face-to-face interaction, which this book argues is critical to the development of interpersonal trust and accurate signal interpretation.

scholarly journals Design and installation of a hot water layer system at the Tehran research reactor

2013 ◽  
Vol 28 (1) ◽  
pp. 18-24
Author(s):  
Sayedeh Mirmohammadi ◽  
Morteza Gharib ◽  
Parnian Ebrahimzadeh ◽  
Reza Amrollahi
Keyword(s):  
Research Reactor ◽  
Cooling Water ◽  
Fission Products ◽  
Water Layer ◽  
Partial Solution ◽  
Time Frame ◽  
Reactor Power ◽  
Hot Water ◽  
Layer System ◽  
Tehran Research Reactor

A hot water layer system (HWLS) is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

Sign in / Sign up

Export Citation Format

Share Document