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 ANALISIS RANCANGAN DASAR SISTEM PGNAA MENGGUNAKAN SUMBER NEUTRON DARI BEAMPORT REAKTOR KARTINI

2017 ◽  
Vol 18 (1) ◽  
pp. 1 ◽  
Author(s):  
Edi Trijono Budisantoso ◽  
Syarip Syarip
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Neutron Source ◽  
Neutron Beam ◽  
Thermal Neutron Flux ◽  
Prompt Gamma ◽  
Basic Design ◽  
Neutron Moderator ◽  
Sample Chamber ◽  
Beam Stopper

ABSTRAK ANALISIS RANCANGAN DASAR SISTEM PGNAA MENGGUNAKAN SUMBER NEUTRON DARI BEAMPORT REAKTOR KARTINI. Telah dilakukan perancangan dasar sistem PGNAA menggunakan salahsatu beamport reaktor Kartini sebagai sumber neutron.  Moderator neutron  ditempatkan pada ujung kolom berkas neutron untuk membuat berkas neutron menjadi termal.  Berkas diarahkan menuju ruang sampel PGNAA dengan menggunakan kolimator yang berfungsi sebagai penyaring berkas neutron sejajar. Pada penggal kolimator yang berpotongan dengan jendela beamport dipasang neutron beam shutter untuk menutup berkas neutron apabila tidak digunakan untuk PGNAA.  Beam stopper dipasang dibelakang ruang sampel PGNAA untuk menangkap berkas neutron yang  lolos. Perhitungan sifat neutronik dilakukan untuk memilih bahan material yang memenuhi syarat fungsi sebagai sub-komponen PGNAA dan menentukan ukuran geometrinya.  Dari hasil perhitungan diperoleh  data bahan yang baik untuk moderator yaitu grafit, bahan kolimator adalah aluminium, bahan beam shutter dan beam stopper adalah komposit boraks-parafin.  Panjang moderator 90 cm, panjang kolimator 173 cm dengan  tetapan kolimasi D/L=0,015, tebal beam shutter dan beam stopper masing-masing 22 cm dan 30 cm.  Dipasang perisai gamma dan perisai neutron untuk menutup berkas neutron keluar dari sela dinding dalam beamport dan didnding luar kolimator. Bahan perisai tersebut dibuat dari komposit boraks parafin 25% berat dan timbal yang masing-masing panjangnya 50 cm dan 30 cm.  Hasil analisis menunjukkan bahwa dari fluks neutron awal pada beamport bagian dalam sebesar 1,5.1012 n/cm2s dapat menghasilkan fluks neutron termal di ruang sampel PGNAA 1,76.108 n/cm2s dengan arus neutron termal 9,29.108 n/s. Nilai fluks neutron termal tersebut memenuhi persyaratan untuk suatu sistem PGNAA yaitu berada pada orde 106 s/d 108 n/cm2s. Kata Kunci : PGNAA, rancangan dasar, prompt-gamma, analisis aktivasi, neutron-termal, beamport reaktor ABSTRACT BASIC DESIGN ANALYSIS OF PGNAA SYSTEM USING NEUTRON SOURCE FROM BEAMPORT OF KARTINI REACTOR. A basic design of PGNAA system using one of reactor beamports of Kartini reactor as a neutron source have been carried out. Neutron moderator is placed at the inner end of beamport column to make thermal neutron beam. A neutron beam directed  to PGNAA counting chamber by using collimator as a filter to make parallel neutron beam.   At  the midle  of collimator intersect with beamport window, neutron beam shutter is mounted to close when not in use for PGNAA.  Beam stopper mounted behind the sample chamber of PGNAA to capture neutron beam that passes from the sample chamber.  Calculation of neutronic properties of materials was done to choose the material that meet the functional requirements of PGNAA and to determine the geometry size.  Based on the calculational results obtained that good material for moderator is graphite, aluminum as beam collimator, and beam shutter or stopper is made from borax-paraffin composite. The moderator length is 90 cm and collimator length  is 173 cm  with collimation constant D / L = 0,015.  Beam shutter and beam stopper thickness are 22 cm and 30 cm respectively. Gamma and neutron shield are  added  surrounding beam colimator to shield  the radiation out from the pitch between collimator and beamport wall.  The shield material made from composite of parrafin 25 w % borax, and lead with the length of 50 cm and 30 cm respectively.   The analysis result shows that from the neutron flux of 1,5.1012 n/cm2s at the inner side of beamport, can generate thermal neutron flux at the PGNAA sample chamber of 1,76.108 n/cm2s with the thermal neutron current of 9,29.108 n/s. This thermal neutron flux meet the requirement for a PGNAA system i.e. in the order of 106 to 108 n/cm2s.   Keywords : PGNAA, basic design, prompt-gamma, activation analysis, thermal neutron, reactor beamport

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 Neutrons production on the IPHI accelerator for the validation of the design of the compact neutron source SONATE

2020 ◽  
Vol 231 ◽  
pp. 01007
Author(s):  
Hoang Ngoc Tran ◽  
Frédéric Ott ◽  
Jacques Darpentigny ◽  
Anthony Marchix ◽  
Alain Letourneau ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Simulations ◽  
Neutron Flux ◽  
Neutron Scattering ◽  
Thermal Neutron ◽  
Neutron Source ◽  
Thermal Neutron Flux ◽  
Gold Foil ◽  
Proton Accelerator ◽  
Good Agreement

We aim at building a compact accelerator-based neutron source (CANS) which would provide a thermal neutron flux on the order of 4x1012 n.s-1.cm-2.sr-1. Such a brilliance would put compact neutron sources on par with existing medium flux neutron research reactors for neutron scattering experiments. We performed the first neutron production tests on the IPHI proton accelerator at Saclay at a proton energy of 3 MeV. The thermal neutron flux were measured using gold foil activation and 3He detectors. The measured flux were compared with GEANT4 Monte Carlo simulations (10.4) in which the whole experimental setup was modelled. There is a good agreement between the experimental measurements and the Monte-Carlo simulations. The available modelling tools will allow us to optimize the whole Target Moderator Reflector assembly together with the neutron scattering spectrometer geometries for the design of the neutron scattering facility SONATE.

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 NEUTRONICS ANALYSIS OF THE FUSION NEUTRON SOURCE TO OBTAIN THE ULTIMATE THERMAL NEUTRON FLUX

2013 ◽  
Vol 36 (1) ◽  
pp. 64-77
Author(s):  
B.V. Kuteev ◽  
◽  
A.V. Zhirkin ◽  
B.K. Chukbar ◽  
V.F. Batyaev ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Neutron Source ◽  
Thermal Neutron Flux ◽  
Fusion Neutron ◽  
Fusion Neutron Source

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.

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