Choosing the optimal conditions for irradiation of specimens in the material testing channel of the VVR-M nuclear reactor

The methodology for choosing the optimal conditions for irradiation of specimens in the material testing channel of the VVR-M research nuclear reactor is presented in the article. The solution to such a problem is necessary to justify the possibility of irradiation of specimens in the material testing channel under given irradiation conditions. In this case, the irradiation conditions include not only the height distribution of the neutron flux density, but also the spectrum of neutrons and the temperature of the specimen in the material testing channel. This approach optimizes the work of VVR-M reactor by placing the maximum possible number of specimens in the material testing channel for irradiation. Also, the optimization of the VVR-M operation involves choosing the location of the research channel in the VVR-M core, where, during the planned irradiation time, the maximum flux density of fast neutrons or neutrons of other energies will be reached, depending on the task. The neutron-physical model of the research nuclear reactor VVR-M in the calculation code SCALE was used for research. The reliability in the determination of neutron-physical characteristics in the VVR-M material testing channel is confirmed by the results of validation carried out at the previous stage of research. It is shown that in order to ensure the necessary accuracy in the determination of the neutron flux parameters in the material testing channel, it is necessary to take into account the fuel burnup, as well as the actual scheme of fuel assemblies rearranging in the VVR-M core for various fuel loads. The results of calculations of important neutronphysical characteristics of the model of a VVR-M nuclear reactor for fuel loading, which is in operation today, on the basis of which it is possible to optimize the choice of the location of the material testing channel in the VVR-M core are presented in the article.

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Determination of the neutron flux density at an unknown reactor power

Annals of Nuclear Energy ◽

10.1016/0306-4549(87)90006-5 ◽

1987 ◽

Vol 14 (12) ◽

pp. 677-680

Author(s):

F. Tomášek

Keyword(s):

Neutron Flux ◽

Flux Density ◽

Reactor Power ◽

Neutron Flux Density

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Determination of the Neutron Flux for the Yankee Rowe Experiment in the Ford Nuclear Reactor

Reactor Dosimetry ◽

10.1520/stp15158s ◽

2009 ◽

pp. 628-628-10

Author(s):

RJ Cacciapouti ◽

L Petrusha

Keyword(s):

Neutron Flux ◽

Nuclear Reactor

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A comparison of some absolute methods of measuring fast neutron flux

Mathematical Proceedings of the Cambridge Philosophical Society ◽

10.1017/s0305004100025810 ◽

1950 ◽

Vol 46 (2) ◽

pp. 339-352 ◽

Cited By ~ 3

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.

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

Volume 4: Codes, Standards, Licensing and Regulatory Issues; Student Paper Competition ◽

10.1115/icone17-76041 ◽

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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