Basic Design Study for the PGAA Facility Components at the CNESTEN´s Triga MArk II reactor

Result code in IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F26722445%3A_____%2F15%3A%230001180" target="_blank" >RIV/26722445:_____/15:#0001180 - isvavai.cz</a>

Result on the web

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DOI - Digital Object Identifier

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Result language

angličtina

Original language name

Basic Design Study for the PGAA Facility Components at the CNESTEN´s Triga MArk II reactor

Original language description

The aim of this study is to provide the basic design of the prompt gamma activation analysis (PGAA) facility components at the TRIGA Mark II reactor operated by CNESTEN in Morocco. The essential need for a PGAA facility is the delivery of a thermal neutron beam with as lowest as possible fast neutrons and gamma background. For this purpose the tangential channel of the reactor was chosen. Compared to the radial channels of the reactor, it provides a better thermal to fast neutron ratio and a lower gamma background. However, the unmodified neutron spectrum is not satisfying for the need of the facility. A common option for modifying the beam parameters to the desired values represent thermal neutron filters. These filters mainly use large single crystals as their main components, which can deliver a high thermal to fast neutron ratio. Additionally, if the filter material is composed of a heavy metal, they suit as a gamma shield as well, and lower, thus, the gamma background. Therefore, the main part of this study is focused on the optimal filter composition to meets the beam requirements for the PGAA facility. The second part of the study aims on the shielding basic design around the modified beam exit, to ensure radiological safety of the operational personal. This comprise the design of the beam shutter and the surrounding shielding blocks which will form the cell, where the sample irradiation and analysis will take place. All the presented calculations were realized using the Monte-Carlo transport code MCNPX 2.7.0. [1] with the ENDF/B-VII data libraries. Additionally, for the filter design approximated modified libraries for thermal neutron scattering on single-crystals were created. All of the radiological calculations use ICRP-74 flux to dose conversion coefficients for the equivalent dose rate calculations.

Czech name

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Czech description

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Type

V_souhrn - Summary research report

CEP classification

JF - Nuclear energy

OECD FORD branch

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Project

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Continuities

N - Vyzkumna aktivita podporovana z neverejnych zdroju

Publication year

2015

Confidentiality

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Number of pages

23

Place of publication

Řež

Publisher/client name

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Version

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