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

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

A comparison of some absolute methods of measuring fast neutron flux

1950 ◽  
Vol 46 (2) ◽  
pp. 339-352 ◽  
Author(s):  
K. W. Allen ◽  
D. L. Livesey ◽  
D. H. Wilkinson
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Cross Sections ◽  
Absolute Measurement ◽  
Fast Neutrons ◽  
Fast Neutron Flux ◽  
Cavendish Laboratory ◽  
The Individual ◽  
Nuclear Processes

The absolute measurement of fast neutron flux presents several difficult problems. Few methods have yet been described in the literature, although the experimental techniques developed by several authors for the detection of fast neutrons (Baldinger, Huber and Staub(7), Barshall and Kanner(9), Amaldi, Bocciarelli, Ferretti and Trabacchi (3), Gray (19), Barshall and Battat(8)) may easily be adapted to this type of measurement. It is, however, most important to have available methods of measuring fast neutron flux to permit the determination of cross-sections for nuclear processes induced by fast neutrons, and several such methods have been developed in the Cavendish Laboratory in recent years. They are the subjects of separate papers (Bretscher and French (13), Kinsey, Cohen and Dainty (21), Allen (l), Allen and Wilkinson (2)). The main purpose of the present paper is to describe the results of experiments carried out to compare these methods in order to test the validity of the assumptions implicit in the individual methods.

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.

Research Reactor VR-1 and Determination of Neutron Spatial Distribution in Its Radial Experimental Channel

2009 ◽  
Author(s):  
Marija Mileticˇ
Keyword(s):  
Spatial Distribution ◽  
Neutron Flux ◽  
Flux Density ◽  
Research Reactor ◽  
Reactor Power ◽  
Activation Detector ◽  
Experimental Channel ◽  
Neutron Flux Density ◽  
Radial Channel

The research reactor VR-1 is operated by Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University (CTU) in Prague. It is a pool-type, light-water reactor, with low enriched uranium. Maximum thermal power is 1kW (equal to 1·108 impulses/second when compared with reactors with higher power). Research on VR-1 reactor is mainly used for the education of university students, preparation and testing of new educational methodologies, investigation of reactor lattice parameters, reactor dynamics study, research in the control equipment field, neutron detector calibration, etc. One of the applications performed by students is the determination of the absolute value of the neutron flux density (also known as Neutron Spatial Distribution) in the radial experimental channel in reactor VR-1. The method used for this measurement is Neutron Activation Analysis. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector which is irradiated in the experimental channel. The activity of the produced radioactive products (radioisotopes) is then measured by means of appropriate counter system (in our case, High Purity Germanium detector). For this measurement totally 34 gold foils were irradiated at different reactor power levels and various positions in radial channel in aim to determine the neutron spatial distribution in radial channel. Interesting results about symmetry, value and dependence on reactor power level of neutron flux density were obtained.

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