Measurement of Fast Neutron, Gamma, and Thermal Neutron Dose Around a Cf-252 Source

2001 ◽  
pp. 539-545
Author(s):  
Somsak Wanwilairat ◽  
Thiraphat Vilaithong ◽  
Vicharn Lorvidhaya ◽  
Rainer Schmidt ◽  
Wilfried Hoffmann
Keyword(s):  
Thermal Neutron ◽  
Fast Neutron ◽  
Neutron Dose

Pulsed thermal neutron source at the fast neutron generator

2009 ◽  
Vol 67 (6) ◽  
pp. 1148-1155 ◽  
Author(s):  
Grzegorz Tracz ◽  
Krzysztof Drozdowicz ◽  
Barbara Gabańska ◽  
Ewa Krynicka
Keyword(s):  
Thermal Neutron ◽  
Fast Neutron ◽  
Neutron Source ◽  
Neutron Generator

Evaluation of thermal neutron dose from Am-Be source using gelatinized boron fricke dosimeter

2019 ◽  
Vol 162 ◽  
pp. 131-135
Author(s):  
E. Aboelezz ◽  
Mahmoud H. Abdelgawad ◽  
S.A. Eman ◽  
G.M. Hassan
Keyword(s):  
Thermal Neutron ◽  
Neutron Dose

scholarly journals Characterization of a 6Li-loaded liquid organic scintillator for fast neutron spectrometry and thermal neutron detection

2013 ◽  
Vol 77 ◽  
pp. 130-138 ◽  
Author(s):  
C.D. Bass ◽  
E.J. Beise ◽  
H. Breuer ◽  
C.R. Heimbach ◽  
T.J. Langford ◽  
...  
Keyword(s):  
Thermal Neutron ◽  
Fast Neutron ◽  
Neutron Detection ◽  
Neutron Spectrometry ◽  
Organic Scintillator ◽  
Liquid Organic Scintillator

The fast neutron energy spectrum and the thermal neutron flux distribution in the experimental channel of the VVR reactor

Atomic Energy ◽  
1962 ◽  
Vol 12 (2) ◽  
pp. 127-132 ◽  
Author(s):  
N. V. Zvonov ◽  
A. I. Mis'kevich ◽  
I. V. Rogozhkin ◽  
V. I. Tereshchenko ◽  
Zh. I. Turkov ◽  
...  
Keyword(s):  
Energy Spectrum ◽  
Neutron Flux ◽  
Thermal Neutron ◽  
Fast Neutron ◽  
Neutron Energy ◽  
Flux Distribution ◽  
Thermal Neutron Flux ◽  
Neutron Energy Spectrum ◽  
Experimental Channel ◽  
Neutron Flux Distribution

Physics Design of Epi-Thermal Neutron Beam for BNCT Based on C-MNSR

2017 ◽  
Author(s):  
Peng Dan ◽  
Wu Xiaobo ◽  
Lu Jin ◽  
Hao Qian ◽  
Hong Jingyan ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Fast Neutron ◽  
Flux Density ◽  
Neutron Beam ◽  
Epithermal Neutron ◽  
Al Alloy ◽  
Neutron Flux Density ◽  
Full Power ◽  
Thermal Neutron Beam

Boron Neutron Capture Therapy (BNCT) is a kind of the targeted therapy with two element. It can kill the cancer cells while the effect on normal cells is very small, and it is suitable for the treatment of the various stage cancer so it will be the ideal radiotherapy for cancer treatment in the future. And Commercial Miniature Neutron Source Reactor (C-MNSR) was designed and constructed by CIAE, which is used for Neutron Activation Analysis (NAA), Training and teaching. The reactor with thermal power 27kW is an under-moderated reactor with pool-tank type, U-AL alloy with High Enriched Uranium (HEU) as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. Design C-MNSR with a epi-thermal neutron beam for BNCT is studied while the conversion from HEU to LEU (Low Enrichment Uranium) (235U percent≤20%) is carried on. As it has the advantages of MNSR safety, economy, easy operation and its application, and it can improve the epi-thermal neutron flux density and meet the requirements of BNCT. The fuel cage of C-MNSR with size of φ230×248mm in the reactor core, there are ton rows of 355lattices are concentrically arranged, the central lattice is reserved for central control rod, and four tie rods are uniformly arranged at the eighth row which link the upper and lower grid plates, the rest 350 fuel lattices are for fuel pins or dummies. The diameter of the fuel meat is 4.3mm, the height is 230mm, with Uranium enrichment is 17%; the diameter of the fuel element is 5.5mm, the height is 248mm. The frame design of the epithermal neutron beam is: Fluental material used as neutron moderation layer with its thickness is 50cm and its density is 2.85g/cm3; Cd with thickness of 0.1cm used as thermal neutron absorption layer, Lead with thickness of 10cm used as gamma ray shielding layer. And the neutron collimator parts is a composition of graphite, Cd and polythene with boron. The total length of the beam is 114.5cm, and the distance from the exit of the beam to the core is 130cm. The results show that the epithermal neutron flux density at the exit is 1.58 × 109n·cm-2·s-1 at full power of 27kW. and the fast neutron density at the exit is 5.45 × 107n · cm-2 · s-1 at full power. Fast neutron dose contamination (Df/ φepi) is 2.88 × 10−11Gy · cm2 · n−1 and gamma dose contamination (Dγ/φepi) 2.18× 10−14 Gy·cm2·n−1.

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