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.

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 Thermal neutron intensity measurement with fission chamber in current, pulse and Campbell modes

2020 ◽  
Vol 231 ◽  
pp. 05009
Author(s):  
Shakir Zeinalov ◽  
Olga Sidorova ◽  
Pavel Sedyshev ◽  
Valery Shvetsov ◽  
Youngseok Lee ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Current Mode ◽  
Fission Chamber ◽  
Design Changes ◽  
Digital Pulse ◽  
Wide Range ◽  
Low Sensitivity ◽  
Digital Pulse Processing

In thermal nuclear reactors, most of the power is generated by thermal neutron induced fission. Therefore, fission chambers with targets that respond directly to slow neutrons are of great interest for thermal neutron flux measurements due to relatively low sensitivity to gamma radiation. However, the extreme conditions associated with experiments at very low cross section demand highly possible thermal neutron flux, leading often to substantial design changes. In this paper we report design of a fission chamber for wide range (from 10 to 1012 n/cm2 sec) measurement of thermal neutron flux. Test experiments were performed at the first beam of IBR2 pulsed reactor using digital pulse processing (DPP) technique with modern waveform digitizers (WFD). The neutron pulses detected by the fission chamber in each burst (5 Hz repetition rate) of the reactor were digitized and recorded to PC memory for further on-line and off-line analysis. New method is suggested to make link between the pulse counting, the current mode and the Campbell technique.

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

A NEW THERMAL NEUTRON FLUX DENSITY STANDARD

1964 ◽  
Vol 42 (8) ◽  
pp. 1593-1604 ◽  
Author(s):  
C. K. Hargrove ◽  
K. W. Geiger
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Flux Density ◽  
Thermal Neutron Flux ◽  
Neutron Detectors ◽  
Coincidence Counting ◽  
Neutron Flux Density ◽  
Graphite Cylinder ◽  
Cylindrical Volume ◽  
Better Than

Six Am241–Be(α, n) neutron sources are arranged around the central circumference of a graphite cylinder 28 cm in diameter by 26.5 cm high. This assembly is surrounded by 10 cm of polyethylene. The graphite cylinder contains an accessible central cavity, 5 cm in diameter by 5 cm high, for the calibration of thermal neutron detectors. The thermal neutron flux density in the cavity is measured by 4πβ–γ coincidence counting of irradiated gold foils. A flux density below cadmium cutoff of nthv0 = 8.70 × 103 cm−2 sec−1 (±1.5%) was found. The uniformity is shown to be constant within the cavity to better than ±0.5% over a cylindrical volume 5 cm in diameter by 2 cm high.

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.

scholarly journals Approximation in Quantum Quadrupole at Juergen Model for Nuclear Reactor Control Rod Blade Based on \ce {Th_xDUO2} Nano Materials

2018 ◽  
Vol 4 (1) ◽  
pp. 7
Author(s):  
Moh. Hardiyanto
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Thermal Neutron Flux ◽  
Nuclear Reactor ◽  
Hadron Collider ◽  
Reactor Control ◽  
Neutron Absorber ◽  
Control Rod ◽  
Chain Reactions ◽  
Nuclear Reactor Control

The functional of a multi purpose research nuclear reactor control rod blade nuclear reactor is stabilized and controlling devices for nuclear chain reactions, the existing of Cerenkov's radiation impact and thermal neutron flux in reactor chamber. This research was conducted in Large Hadron Collider (LHC) - Muon Hadron Division at CERN, Lyon - France under International Research between Canadian Deuterium Uranium (CANDU) - Nuclear Reactor and Betha Group Section for sub-particles for nanomaterial. Using Juergen Model with quantum states approaching and testing by Muon-Hadron Stirrer equipment had determined the \ce {Th_xDUO2} derivatives materials. This material shown the strength of thermal neutron flux absorbed about 2.56 × 10⁵ − 1.94 × 10⁶ Ci/mm, the value of Electrical Conductivity is 26.62 − 29.98 in 800° - 890° C temperature, however at 2.1 × 10⁵ Ci/mm thermal neutron flux condition is 29.44 − 37.88 in IAEA standard. At 450 tesla magnetic field and 2.1 × 10⁵ Ci/mm thermal neutron absorber, the crystalline structure reduction is 6.88% until 10.95% for 25 years period in 45.7 megawatts with \ce {UO2} more enrichment and \ce {Pu2O} also \ce {Th2O_y} nuclear fuel element matrix.

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

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