Neutron flux parameters at irradiation positions in the new research reactor FRM-II

2006 ◽  
Vol 564 (2) ◽  
pp. 641-644 ◽  
Author(s):  
Xilei Lin ◽  
Richard Henkelmann ◽  
Andreas Türler ◽  
Heiko Gerstenberg ◽  
Frans De Corte
Keyword(s):  
Neutron Flux ◽  
Research Reactor ◽  
Neutron Flux Parameters ◽  
New Research

scholarly journals Determination of neutron flux parameters in PUSPATI TRIGA Mark II Research Reactor, Malaysia

2012 ◽  
Vol 296 (3) ◽  
pp. 1231-1237 ◽  
Author(s):  
I. A. Alnour ◽  
H. Wagiran ◽  
N. Ibrahim ◽  
S. Hamzah ◽  
B. S. Wee ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Research Reactor ◽  
Neutron Flux Parameters ◽  
Triga Mark Ii

Neutron flux parameters for k0-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

2011 ◽  
Vol 46 (2) ◽  
pp. 219-223 ◽  
Author(s):  
A.R. Yavar ◽  
S. Sarmani ◽  
A.K. Wood ◽  
S.M. Fadzil ◽  
Z. Masood ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Research Reactor ◽  
Neutron Flux Parameters

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 A Design of Boron Neutron Capture Therapy for Cancer Treatment in Indonesia

2016 ◽  
Vol 1 (1) ◽  
pp. 1
Author(s):  
Yohannes Sardjono ◽  
Susilo Widodo ◽  
Irhas Irhas ◽  
Hilmi Tantawy
Keyword(s):  
Neutron Flux ◽  
Boron Neutron Capture Therapy ◽  
Neutron Capture ◽  
Neutron Generator ◽  
Epithermal Neutron ◽  
Research Reactor ◽  
Neutron Capture Therapy ◽  
Epithermal Neutron Flux ◽  
Boron Neutron Capture ◽  
Compact Neutron Generator

Boron Neutron Capture Therapy (BNCT) is an advanced form of radiotherapy technique that is potentially superior to all conventional techniques for cancer treatment, as it is targeted at killing individual cancerous cells with minimal damage to surrounding healthy cells. After decades of development, BNCT has reached clinical-trial stages in several countries, mainly for treating challenging cancers such as malignant brain tumors. The Indonesian consortium of BNCT already developed of the design BNCT for many cases of type cancers using many neutron sources. The main objective of the Indonesian consortium BNCT are the development of BNCT technology package which consists of a non nuclear reactor neutron source based on cyclotron and compact neutron generator technique, advanced boron-carrying pharmaceutical, and user-friendly treatment platform with automatic operation and feedback system as well as commercialization of the BNCT though franchised network of BNCT clinics worldwide. The Indonesian consortium BNCT will offering to participate in Boron carrier pharmaceuticals development and testing, development of cyclotron and compact neutron generators and provision of neutrons from the 100 kW Kartini Research Reactor to guide and to validate compact neutron generator development. Studies were carried out to design a collimator which results in epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle 5 (MCNP5) codes. Reactor within 100 kW of output thermal power was used as the neutron source. The design criteria were based on the IAEA’s recommendation. All materials used were varied in size, according to the value of mean free path for each. Monte Carlo simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 1,5 cm thick of Bi as "-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 3-5 cm varied aperture size, epithermal neutron beam with minimum flux of 7,8 x 108 n.cm-2.s-1, maximum fast neutron and "-ray components of, respectively, 1,9 x 10-13 Gy.cm2.n-1 and 1,8 x 10-13 Gy.cm2.n-1, maximum thermal neutron per epithermal neutron ratio of 0,009, and beam minimum directionality of 0,72, could be produced. The beam did not fully pass the IAEA’s criteria, since the epithermal neutron flux was still below the recommended value, 1,0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeded 5 x 108 n.cm-2.s-1. When this collimator was surrounded by 8 cm thick of graphite, the characteristics of the beam became better that it passed all IAEA’s criteria with epithermal neutron flux up to 1,7 x 109 n.cm-2.s-1. it is still feasible for BNCT in vivo experiment and study of many cases cancer type i.e.; liver and lung curcinoma. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Sodium boroncaptate (BSH) was used as in this research. BSH had effected in liver for radiation quality factor as 0.8 in health tissue and 2.5 in cancer tissue. Modelling organ and source used liver organ who contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 $g/g cancer. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Given the advantages of low density owned by lungs, hence BNCT is a solid option that can be utilized to eradicate the cell cancer in lungs. Modelling organ and neutron source for lung carcinoma was used Compact Neutron Generator (CNG) by deuterium-tritium which was used is boronophenylalanine (BPA). The concentration of boron-10 compound was varied in the study; i.e. the variations were 20; 25; 30; 35; 40 and 45 μg.g-1 cancer tissues. Ideally, the primary dose which is solemnly expected to contribute in the therapy is alpha dose, but the secondary dose; i.e. neutron scattering dose, proton dose and gamma dose that are caused due to the interaction of thermal neutron with the spectra of tissue can not be simply omitted. Thus, the desired output of MCNPX; i.e. tally, were thermal and epithermal neutron flux, neutron and photon dose. The liver study variation of boron concentration result dose rate to every variation were0,042; 0,050; 0,058; 0,067; 0,074; 0,082; 0,085 Gy/sec. Irradiation time who need to every concentration were 1194,687 sec (19 min 54 sec);999,645 sec (16 min 39 sec); 858,746 sec (14 min 19 sec); 743,810 sec (12 min 24 sec); 675,156 sec (11 min 15 sec); 608,480 sec (10 min 8 sec); 585,807sec (9 min 45 sec). The lung carcinoma study variations of boron-10 concentration in tissue resulted in the dose rate of each variables respectively were 0.003145, 0.003657, 0.00359, 0.00385, 0.00438 and 0.00476 Gy.sec-1 . The irradiated time needed for therapy for each variables respectively were 375.34, 357.55, 287.58, 284.95, 237.84 and 219.84 minutes.

Measurement of neutron flux distribution in the irradiation channel in the Ghana Research Reactor-1 using Monte Carlo method

2011 ◽  
Vol 53 (2) ◽  
pp. 189-194 ◽  
Author(s):  
R.G. Abrefah ◽  
S. Anim-Sampong ◽  
B.J.B. Nyarko ◽  
E.H.K. Akaho ◽  
R.B.M. Sogbadji
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Neutron Flux ◽  
Flux Distribution ◽  
Research Reactor ◽  
Irradiation Channel ◽  
Neutron Flux Distribution

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.

scholarly journals Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code

2018 ◽  
Vol 11 (3) ◽  
pp. 255-261 ◽  
Author(s):  
Jamal Al Zain ◽  
O. El Hajjaji ◽  
T. El Bardouni ◽  
H. Boukhal
Keyword(s):  
Neutron Flux ◽  
Research Reactor ◽  
Safety Parameters
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