Neutronic Analysis for Tehran Research Reactor Upgrading

2002 ◽  
Author(s):  
Ebrahim Afshar ◽  
Alireza Shahidi
Keyword(s):  
Research Reactor ◽  
Three Dimensional ◽  
Operating Conditions ◽  
Monte Carlo Code ◽  
Radioisotope Production ◽  
The Core ◽  
Transport Code ◽  
Tehran Research Reactor ◽  
Fuel Elements ◽  
Good Agreement

This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of Tehran Research Reactor (TRR) from 5 to 10 MW (th), keeping the same core configuration and with minimum changes in the primary cooling circuit. The main aim of TRR upgrade is to increase the volume of radioisotope production. The neutronic analysis was carried out for a fresh core with 22 Standard Fuel Elements (SFE) under normal operating conditions. Two different calculational lines were used to simulate the neutronic behavior in the core and perform the necessary neutronic calculations. First, combination of cell calculation transport code WIMS-D/4 [1] and three dimensional core calculation diffusion code CITATION [2] were used to and next a Monte Carlo code MCNP-4B [3] together with point depletion code ORIGEN-2 [4] were used. The results obtained show good agreement between these two different schemes.

scholarly journals Core parametric study for enhancing the radioisotope production in the IEA-R1 research reactor

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Giovanni Laranjo Stefani ◽  
Frederico Antônio Genezini ◽  
Thiago Augusto Santos ◽  
João Manoel de Losada Moreira
Keyword(s):  
Neutron Flux ◽  
Parametric Study ◽  
Research Reactor ◽  
Reactor Core ◽  
Radioisotope Production ◽  
Spent Fuel ◽  
The Core ◽  
Safety Parameters ◽  
Fuel Elements ◽  
Average Neutron

In this work a parametric study was carried to increase the production of radioisotopes in the IEA-R1 research reactor. The changes proposed to implement in the IEA-R1 reactor core were the substitution of graphite reflectors by beryllium reflectors, the removal of 4 fuel elements to reduce the core size and make available 4 additional locations to be occupied by radioisotope irradiation devices. The key variable analyzed is the thermal neutron flux in the irradiation devices.  The proposed configuration with 20 fuel elements in an approximately cylindrical geometry provided higher average neutron flux (average increment of 12.9 %) allowing higher radioisotope production capability. In addition, it provided 4 more positions to install  irradiation devices which allow a larger number of simultaneous irradiations practically doubling the capacity of radioisotope production in the IEA-R1 reactor. The insertion of Be reflector elements in the core has to be studied carefully since it tends to promote strong neutron flux redistribution in the core. A verification of design and safety parameters of the proposed  core was carried out. The annual fuel consumption will increase about 17 % and more storage space for spent fuel will be required.   

Numerical Calculation of Pressure Fluctuations in the Volute of a Centrifugal Fan

2005 ◽  
Vol 128 (2) ◽  
pp. 359-369 ◽  
Author(s):  
Rafael Ballesteros-Tajadura ◽  
Sandra Velarde-Suárez ◽  
Juan Pablo Hurtado-Cruz ◽  
Carlos Santolaria-Morros
Keyword(s):  
Fluid Dynamics ◽  
Pressure Fluctuations ◽  
Three Dimensional ◽  
Power Spectra ◽  
Operating Conditions ◽  
Experimental Results ◽  
Time Dependent ◽  
Centrifugal Fan ◽  
Wall Pressure Fluctuations ◽  
Good Agreement

In this work, a numerical model has been applied in order to obtain the wall pressure fluctuations at the volute of an industrial centrifugal fan. The numerical results have been compared to experimental results obtained in the same machine. A three-dimensional numerical simulation of the complete unsteady flow on the whole impeller-volute configuration has been carried out using the computational fluid dynamics code FLUENT®. This code has been employed to calculate the time-dependent pressure both in the impeller and in the volute. In this way, the pressure fluctuations in some locations over the volute wall have been obtained. The power spectra of these fluctuations have been obtained, showing an important peak at the blade passing frequency. The amplitude of this peak presents the highest values near the volute tongue, but the spatial pattern over the volute extension is different depending on the operating conditions. A good agreement has been found between the numerical and the experimental results.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

Collimated neutron beam design for TRR thermal column

2021 ◽  
Vol 16 (12) ◽  
pp. P12023
Author(s):  
M.J. Mozafari Vanani ◽  
Y. Kasesaz ◽  
M. Hosseinipanah ◽  
A. Akhound
Keyword(s):  
Monte Carlo ◽  
Neutron Beam ◽  
Research Reactor ◽  
Neutron Radiography ◽  
Monte Carlo Code ◽  
Thermal Neutrons ◽  
Beam Design ◽  
Thermal Column ◽  
Tehran Research Reactor ◽  
Thermal Neutron Beam

Abstract Tehran Research Reactor (TRR) is the main neutron source in Iran which can be used for different applications of neutrons such as neutron radiography and neutron therapy. TRR has a thermal column which can provide high intensity flux of thermal neutrons for users. The aim of this study is to design a neutron collimator for TRR thermal column to produce parallel neutron beam with suitable intensity of thermal neutrons. To achieve this goal, Monte Carlo code of MCNX has been used to evaluate different configurations, geometries and materials of neutron collimator. The results show that the final selected configuration can provide a uniform thermal neutron beam with a flux of 1.21E+13 (cm-2·s-1) which is suitable for many different neutron applications.

scholarly journals Thermally Induced Deformations in Nuclear Fuel Elements

2021 ◽  
Author(s):  
Haihui (Stella) Yang
Keyword(s):  
Fuel Element ◽  
Transfer Factor ◽  
Numerical Solutions ◽  
Three Dimensional ◽  
Beam Theory ◽  
Operating Conditions ◽  
Operational Parameters ◽  
Thermally Induced ◽  
Sensitivity Studies ◽  
Fuel Elements

Nonlinear three-dimensional multibody surface-surface contacts, thermally induced deformations, and the curvature transfer factor in CANDU fuel elements are investigated using the finite element method in this thesis. ANSYS is selected to obtain numerical solutions for CANDU fuel elements under several operating conditions. In the ANSYS models, the 20-node structural elements (SOLID186) are employed to mesch individual solids; the surface-to surface contact pairs (TARGE170 and CONTA174) are used to handle contacts between solids. Sensitivity studies on the curvature transfer factor are conducted for several key operational parameters. If there is full radial contact between the pellets and the sheath, a CANDU fuel element may be considered as a composite beam because of the large length-to-diameter ratio. The Timoshenko beam theory is used in conjunction with a three-node mean element to explore the thermal deformation behaviours of a fuel element. A program written in MATLAB is much more efficient compared with the ANSYS solutions.

A New Low Enrichment Uranium Core Design of MNSR

2017 ◽  
Author(s):  
Wang Mengjiao ◽  
Li Yiguo ◽  
Wu Xiaobo ◽  
Peng Dan ◽  
Hong Jingyan ◽  
...  
Keyword(s):  
Optimization Design ◽  
Research Reactor ◽  
Nuclear Safety ◽  
Fuel Pellet ◽  
Design Parameters ◽  
Control Rod ◽  
The Core ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Better Than

The Miniature Neutron Source Reactor (MNSR) is a low-power research reactor, which uses 90% high enriched uranium (HEU) fuel. However, due to the nuclear safety risk, and according to the principle of nuclear non-proliferation, MNSR must be gradually converted from HEU to low enriched uranium (LEU), which means the LEU fuel with U-235 enrichment less than 20% should be used. The prototype MNSR of China Institute of Atomic Energy has completed the transformation, but other commercial MNSRs have not finished, which is different with the prototype in the application and structure. Therefore, using MCNP code to simulate, calculate and optimization design LEU core has been done in this issue. Firstly, UO2 with U-235 enrichment of 12.5% was selected as the fuel pellet of LEU core, keeping the rest of the core unchanged. The Φ, excess reactivity and the worth of the central control rod are calculated and analyzed. The results show that the commercial MNSR of LEU conversion is feasible. Secondly, in this paper, through changing the fuel elements and the arrangement method, the new low enriched uranium (NLEU) core was designed to improve Φ/P ratio of the core, the proportion of thermal neutrons and the worth of the control rod. UO2 with U-235 enrichment of 19.75% was selected as the fuel pellet of the NLEU, NLEU not only meets the design parameters, but in many parameters, NLEU is better than LEU. The fuel element quantity is reduced by 43%, from original 344 to 196; reducing the amount of U-235 loading; improving the Φ/P ratio and the thermal neutron fraction is increased. The results show that the NLEU optimizes some parameters, simplifies the core structure, saves the construction cost, improves the nuclear safety and is more suitable for the application of MNSR.

A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification

2018 ◽  
Author(s):  
Wankui Yang ◽  
Baoxin Yuan ◽  
Songbao Zhang ◽  
Haibing Guo ◽  
Yaoguang Liu ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Research Reactor ◽  
Neutron Transport ◽  
Reactor Core ◽  
Deep Penetration ◽  
Monte Carlo Code ◽  
Deterministic Theory ◽  
Continuous Energy ◽  
Transport Calculation ◽  
Good Agreement

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing keff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But keff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the keff difference between these two methods remains. Even though the keff results in this benchmark are less than keff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

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