Neutron flux from a 14-MeV neutron generator with tungsten filter for research in NDA methods for nuclear safeguards and security

2009 ◽  
Author(s):  
H. Rennhofer ◽  
B. Pedersen ◽  
J.-M. Crochemore ◽  
K. Baharuth-Ram
Keyword(s):  
Neutron Flux ◽  
Neutron Generator ◽  
Nuclear Safeguards

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.

Neutron flux distribution from a 14 MeV neutron generator

1967 ◽  
Vol 50 (1) ◽  
pp. 141-146 ◽  
Author(s):  
H.F. Priest ◽  
F.C. Burns ◽  
G.L. Priest
Keyword(s):  
Neutron Flux ◽  
Neutron Generator ◽  
Flux Distribution ◽  
Neutron Flux Distribution

High spatial resolution fiber based miniature detector for neutron flux mapping in a subcritical assembly and neutron generator

2017 ◽  
Vol 19 (3-4) ◽  
pp. 169-175 ◽  
Author(s):  
Saroj Bishnoi ◽  
P.S. Sarkar ◽  
Mayank Shukla ◽  
Nirmal Ray ◽  
Tarun Patel ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Spatial Resolution ◽  
Neutron Generator ◽  
High Spatial Resolution ◽  
Flux Mapping

scholarly journals Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

2018 ◽  
Vol 177 ◽  
pp. 02003 ◽  
Author(s):  
Aya Hamdy Hegazy ◽  
V.R. Skoy ◽  
K. Hossny
Keyword(s):  
Neutron Flux ◽  
Neutron Beam ◽  
Gamma Rays ◽  
Neutron Generator ◽  
Nuclear Research ◽  
Experimental Results ◽  
Fast Neutrons ◽  
Gamma Line ◽  
Total Neutron ◽  
Previous Conclusion

Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

Angular Distribution of Neutron Flux around the Tritium Target of 14 MeV Neutron Generator

2012 ◽  
Vol 81 (10) ◽  
pp. 104006 ◽  
Author(s):  
Ranjita Mandal ◽  
Genu Radhu Pansare ◽  
Debashish Sengupta ◽  
Vasant Nageshrao Bhoraskar
Keyword(s):  
Angular Distribution ◽  
Neutron Flux ◽  
Neutron Generator ◽  
Tritium Target

Determination of the Neutron Flux of the NG-12I Neutron Generator

2005 ◽  
Vol 48 (8) ◽  
pp. 828-833 ◽  
Author(s):  
V. I. Litvin ◽  
Ya. Z. Kandiev ◽  
G. N. Malyshkin ◽  
E. A. Kashaeva ◽  
S. I. Samarin ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Neutron Generator

scholarly journals The neutron production rate measurement of an indigenously developed compact D-D neutron generator

2013 ◽  
Vol 28 (4) ◽  
pp. 422-426
Author(s):  
Basanta Das ◽  
Anurag Shyam ◽  
Rashmita Das ◽  
Durga Rao
Keyword(s):  
Neutron Flux ◽  
Neutron Generator ◽  
Flux Measurement ◽  
Ion Source ◽  
Equivalent Rate ◽  
Titanium Target ◽  
Compact Neutron Generator ◽  
Rate Conversion ◽  
Deuterium Gas ◽  
Technology Industries

One electrostatic accelerator based compact neutron generator was developed. The deuterium ions generated by the ion source were accelerated by one accelerating gap after the extraction from the ion source and bombarded to a target. Two different types of targets, the drive - in titanium target and the deuteriated titanium target were used. The neutron generator was operated at the ion source discharge potential at +Ve 1 kV that generates the deuterium ion current of 200 mA at the target while accelerated through a negative potential of 80 kV in the vacuum at 1.3?10-2 Pa filled with deuterium gas. A comparative study for the neutron yield with both the targets was carried out. The neutron flux measurement was done by the bubble detectors purchased from Bubble Technology Industries. The number of bubbles formed in the detector is the direct measurement of the total energy deposited in the detector. By counting the number of bubbles the total dose was estimated. With the help of the ICRP-74 neutron flux to dose equivalent rate conversion factors and the solid angle covered by the detector, the total neutron flux was calculated. In this presentation the operation of the generator, neutron detection by bubble detector and estimation of neutron flux has been discussed.

Simulation Study on the Moderator in PGNAA-Based Online Coal Measurement System

2014 ◽  
Vol 675-677 ◽  
pp. 1316-1320
Author(s):  
Qing Shan ◽  
Qun Li ◽  
Can Cheng ◽  
Wen Bao Jia ◽  
Da Qian Hei ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Measurement System ◽  
Neutron Generator ◽  
Prompt Gamma ◽  
Energy Neutron ◽  
Preliminary Study ◽  
Simulation Results ◽  
Flux Increase ◽  
Gamma Neutron Activation

An online coal measurement system, which is based on the prompt gamma neutron activation analysis (PGNAA) technology and uses a D-T neutron generator, is studied in this work. To improve the analytical precision of the element in coal, the original moderator is optimized. The standards of the optimization are (1) the neutron flux increase 50% after moderation; (2) the proportion of thermal and middle-energy neutron and fast neutron is nearly equal to 1:1. Using Monte Carlo method, the moderator has been optimized by adding the uranium238 to the original moderator, which consists of high density polyethylene (HDPE). The simulation results show that the optimization standards can be basically satisfied when the thickness of the uranium238 and HDPE are 4cm and 1cm respectively. On this basis, preliminary study on introducing the channel to moderator is carried out. The preliminary results show that the channel can improve the total neutron flux. But also, it will decrease the proportion of thermal and middle-energy neutron and fast neutron.

scholarly journals Application of fluorine-based threshold activation detector for neutron flux calculation from D-T neutron generator

2020 ◽  
Vol 225 ◽  
pp. 02004
Author(s):  
Paweł Sibczyński ◽  
Andrzej Brosławski ◽  
Szymon Burakowski ◽  
Arkadiusz Chłopik ◽  
Marek Dryll ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Neutron Generator ◽  
Distribution Curve ◽  
Fast Neutron Flux ◽  
Analysis Method ◽  
Activation Detector ◽  
Scintillation Crystal ◽  
Activation Reaction ◽  
Transport Code ◽  
Flux Estimation

In this paper we propose a method of fast neutron flux estimation from a pulsed D-T neutron generator with application of single CaF2 scintillation crystal. The analysis method relies on 19F(n, α)16N threshold activation reaction having neutron energy threshold at 1.6 MeV. As a result, the 16N undergo β− decay with half-life of 7.1 s, emitting β particles with endpoint up to 10.4 MeV in the scintillator medium. Integration of the β distribution curve, preceded by calculation of (n, α) rate on F with Monte Carlo N-Particle Transport Code v6 (MCNP6) for fixed geometry, allows to estimate the neutron flux in 4π per second within few minutes.

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