Analytical estimation of the central reflector impact on thermal and fast neutron flux density in research reactors

2016 ◽  
Author(s):  
D. V. Gvozdyakov ◽  
A. V. Kuzmin ◽  
A. O. Tanishev ◽  
S. A. Shvab
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Flux Density ◽  
Fast Neutron Flux ◽  
Research Reactors ◽  
Neutron Flux Density ◽  
Analytical Estimation

Monitoring the Fast-Neutron Flux Density and Fluence in a RBMK Core using a Threshold Fission Chamber in a Screen-Absorber

Atomic Energy ◽  
2005 ◽  
Vol 98 (4) ◽  
pp. 249-255 ◽  
Author(s):  
Yu. A. Kashchuk ◽  
V. G. Shevchenko ◽  
V. V. Frunze ◽  
A. V. Krasil’nikov
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Flux Density ◽  
Fast Neutron Flux ◽  
Fission Chamber ◽  
Neutron Flux Density ◽  
Rbmk Core

Determination of the fast-neutron flux density in a reactor of the Kola Atomic Power Plant

1982 ◽  
Vol 52 (6) ◽  
pp. 436-437
Author(s):  
Kh. Ya. Bondars ◽  
E. I. Ignatenko ◽  
A. A. Lapenas ◽  
V. I. Lobov ◽  
S. S. Lomakin ◽  
...  
Keyword(s):  
Power Plant ◽  
Neutron Flux ◽  
Fast Neutron ◽  
Flux Density ◽  
Atomic Power Plant ◽  
Fast Neutron Flux ◽  
Neutron Flux Density

scholarly journals Analytical Estimation of the Central Reflector Impact on Thermal Neutron Flux Density in Research Reactors

2015 ◽  
Vol 37 ◽  
pp. 01024 ◽  
Author(s):  
Dmitry V. Gvozdyakov ◽  
Anton O. Tanishev ◽  
Svetlana A. Shvab ◽  
Vladimir N. Martyshev ◽  
Arian V. Kuzmin
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Flux Density ◽  
Thermal Neutron Flux ◽  
Research Reactors ◽  
Neutron Flux Density ◽  
Analytical Estimation

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.

Comparison of Neutron Flux Density in Carbon Prism filled with NaCl and Air

2020 ◽  
Author(s):  
Kamil Stevanka ◽  
Dusan Kral ◽  
Ondrej Stastny ◽  
Robert Holomb ◽  
Karel Katovsky ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Flux Density ◽  
Neutron Flux Density

KNVK vacuum fission chambers for measuring neutron flux density

1987 ◽  
Vol 62 (3) ◽  
pp. 232-237
Author(s):  
E. K. Malyshev ◽  
S. V. Chuklyaev ◽  
O. I. Shchetinin
Keyword(s):  
Neutron Flux ◽  
Flux Density ◽  
Neutron Flux Density ◽  
Fission Chambers
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