A methodology to enhance thermal neutron flux in Tehran Research Reactor core for domestic fuel test purposes

2021 ◽  
Vol 136 ◽  
pp. 103726
Author(s):  
S. Safaei Arshi ◽  
A. Jozvaziri ◽  
S.M. Mirvakili ◽  
M. Keyvani
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Reactor Core ◽  
Tehran Research Reactor ◽  
Domestic Fuel

scholarly journals DESIGN OF IRRADIATION FACILITIES AT CENTRAL IRRADIATION POSITION OF PLATE TYPE RESEARCH REACTOR BANDUNG

2020 ◽  
Vol 22 (1) ◽  
pp. 1
Author(s):  
Epung Saepul Bahrum ◽  
Wawan Handiaga ◽  
Yudi Setiadi ◽  
Henky Wibowo ◽  
Prasetyo Basuki ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Flux Distribution ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Reactor Core ◽  
Irradiation Position ◽  
Neutron Flux Distribution ◽  
Flux Profile ◽  
Plate Type

One of the results from Plate Type Research Reactor Bandung (PTRRB) research program is PTRRB core design. Previous study on PTRRB has not calculated neutron flux distribution at its central irradiation position (CIP). Distribution of neutron flux at CIP is of high importance especially in radioisotope production. In this study, CIP was modeled as a stack of four to five aluminum tubes (AT), each filled by four aluminum irradiation capsules (AIC). Considering AIC dimension and geometry, there are three possibilities of AT configuration. For irradiation sample, 1.45 gr of molybdenum (Mo) was put into AIC. Neutron flux distribution at Mo sample was calculated using TRIGA MCNP and MCNP software. The calculation was simulated at condition when fresh fuel is loaded into reactor core. Analyses of excess reactivity show that, after installing irradiation AT and Mo sample was put into each configuration, the excess reactivity is less than 10.9 %. The highest calculated thermal neutron flux at Mo sample is 5.08×1013 n/cm2.s at configuration 1. Meanwhile, the highest total neutron flux at Mo sample is located at capsule no. II and III. Thermal neutron flux profile is the same for all configurations. This result will be used as a basic data for PTRRB utilization.Keywords: Central Irradiation Position, Neutron Flux Distribution, MCNP, PTRRB

The Advanced Neutron Source

1989 ◽  
Vol 166 ◽  
Author(s):  
John B. Hayter
Keyword(s):  
Steady State ◽  
Neutron Flux ◽  
Thermal Neutron ◽  
Neutron Source ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Experimental Facility ◽  
Oak Ridge ◽  
State Research

ABSTRACTThe Advanced Neutron Source (ANS) is a new user experimental facility planned to be operational at Oak Ridge in the late 1990's. The centerpiece of the ANS will be a steady-state research reactor of unprecedented thermal neutron flux (φth ≍ 9·1019 m−2·s−1) accompanied by extensive and comprehensive equipment and facilities for neutron-based research.

scholarly journals Toward for production of Molybdenum-99 by irradiation of MoO3 target in a neutron flux

2018 ◽  
Vol 17 (4) ◽  
pp. 567-572 ◽  
Author(s):  
Abdessamad Didi ◽  
Ahmed Dadouch ◽  
Otman Jai ◽  
Fatima Zahra Bouhali
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Nuclear Energy ◽  
Thermal Neutron Flux ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Radiation Treatment ◽  
Medical Science ◽  
Fortran 90 ◽  
Triga Mark Ii

Background: Molybdenum- 99 is a parent isotope of Technetium-99m (99mTc) as intermediate to diagnosis and radiation treatment. This production is made according to irradiation of Molybdenum-98 with thermal neutrons. The cycle comprises a complex of MoO3 or the percentage of Mo-98 is 24.13%, this compressed mixture in an irradiation capsule of aluminum; the latter is disposed in the central thimble of a nuclear reactor so that the thermal neutron flux is at a maximum in order to generate 98Mo to 99Mo by nuclear reaction of (n, γ) where the crosssection of molybdenum is (0.13 ± 0.0013) barn.Method: The purpose of this study is to validate of MoO3 target that will be used for the production of Molybdenum-99 in the central thimble irradiation position of the TRIGA Mark II research reactor at CNESTEN (Morocco National Center for Nuclear Energy, Sciences and Techniques), The thermal neutron flux used for activity calculation of Mo-99 used reactor of research TRIGA Mark-II is 3.1 1011 n/cm²s at 250 KW, 2.4E+13 n/cm2s at 1.1 MW and 3.01 1013n/cm²s at 2MW, we are using Fortran-90 code for calculate the activity.Result: The result’s finding was validated by other studies.Bangladesh Journal of Medical Science Vol.17(4) 2018 p.567-572

scholarly journals Design and Construction of a Preamplifier for Research Reactor Control System Using Russia’s Neutron Detectors

2016 ◽  
Vol 6 (3) ◽  
pp. 8-15
Author(s):  
Dinh Hai Trinh ◽  
Van Tai Vo ◽  
Van Diep Le ◽  
Nhi Dien Nguyen
Keyword(s):  
Control System ◽  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Neutron Detectors ◽  
Reactor Control ◽  
Wide Range ◽  
Reactor Control System ◽  
Design And Construction

This paper presents the design and construction of a preamplifier device for Research Reactor Control System, using Russia’s Neutron Detectors of ionization and fission chambers. In this work, the preamplifier device which consists of a wide range Current to Frequency Converter block used with a compensation ionization chamber type KNK-3 to measure the thermal neutron flux in the range of 1x106 ¸ 1x1011 n/cm2.s, a Pulse Preamplifier block used with a fission chamber type KNK-15 to measure the thermal neutron flux in the range of 1x100 ¸ 1x106 n/cm2.s, and a Power Supply block, was designed and tested in different conditions in the laboratory and at Dalat Nuclear Research Reactor (DNRR). Obtained results show that, the above blocks have almost design specifications as equivalent or better in comparison with the same function blocks of the DNRR’s Control System which were designed by the former Soviet Union. They also meet the utilization requirements as well as the experimental and training purposes.

scholarly journals Design of an irradiation rig using screen method for silicon transmutation doping at the Dalat research reactor

2019 ◽  
Vol 9 (1) ◽  
pp. 1-8
Author(s):  
Huy Pham Quang ◽  
Dien Nguyen Nhi ◽  
Cuong Nguyen Kien ◽  
Duong Tran Quoc ◽  
Dang Vo Doan Hai ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Flux Ratio ◽  
Nuclear Research ◽  
Research Centre ◽  
Screen Method ◽  
Transmutation Doping ◽  
Design And Testing

The neutron transmutation doping of silicon (NTD-Si) at research reactors has beensuccessfully implemented in many countries to produce high-quality semiconductors. In the late 1980s, NTD-Si has been tested at the Dalat Nuclear Research Reactor (DNRR) but the results have been limited. Therefore, the design and testing of an irradiation rig for NTD-Si at the DNRR are necessary to have a better understanding in order to apply the NTD-Si in a new research reactor of the Research Centre for Nuclear Science and Technology (RCNEST), which has planned to be built in Viet Nam. This paper presents the design and testing of a new irradiation rig using screen method for testing NTD-Si at the DNRR. The important parameters in the rig such as neutron spectrum and thermal neutron flux distribution were determined by both calculation using MCNP5 computer code and experiment. The aluminum ingots, which have similar neutronic characteristics with silicon ingots, were irradiated in the rig to verify the appropriate design. The uniformity of thermal neutron flux in the rig is less than 5% in axial and 2% in radial directions, respectively. However, the thermal/fast flux ratio of the irradiation rig is 4.38/1 would affect target resistivity of testing Silicon ingots after irradiation.

Thermal Neutron Flux Distributions in Metal-Hydrogenous Shields

1960 ◽  
Vol 8 (3) ◽  
pp. 260-269 ◽  
Author(s):  
D. C. Anderson ◽  
K. Shure
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux
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