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Effect of Boron Impurity and Graphite Thermal Neutron Scattering on Criticality Calculation of Indonesian Experimental Power Reactor (Suwoto, H. Adrial, W. Luthfi, T. Setiadipura, Zuhair)

Suwoto, AW and Hery Adrial, HA and Wahid Luthfi, WL and Topan Setiadiapura, TS and Zuhair, Z (2019) Effect of Boron Impurity and Graphite Thermal Neutron Scattering on Criticality Calculation of Indonesian Experimental Power Reactor (Suwoto, H. Adrial, W. Luthfi, T. Setiadipura, Zuhair). In: ICoNetS 2019 Universitas Andalas Padang Sumatera Barat, 18 September 2019, Universitas ANDALAS Fakultas Teknik Padang Sumatera Barat.

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Abstract

Abstract. The structural materials of Indonesian Experimental Reactor (RDE) is made from graphite that dominate material used on core structure. So that graphite material is very important role, both as core structure material, reflector and also as a moderator and fuel layer and fuel matrix. In thermal neutron energy range, the neutron scattering collision in moderator material such as graphite influences the neutron cross-section and the resulting energy distribution, so that neutrons will get an increase in energy for excitation in the material. Due to high neutron absorption cross section, boron and its compounds find extensive application in the nuclear industry. In actually it is very difficult to obtain pure uranium or thorium dioxide without any other substance. Usually uranium dioxide or thorium kernel always has impurity material like boron. Boron is one of the materials that has strong neutron absorber, specially for Boron-10.The research starting from modeling of kernel TRISO coated fuel particle, spherical pebble fuel and full core modeling by involving multiple heterogeneity calculations. Boron impurities in the TRISO kernel coated fuel particles was carried out with 27 data varied concentration of boron are 0ppm, 1ppm, 2ppm, 3ppm, 4ppm, 5ppm, 6ppm, 7ppm, 8ppm, 9ppm, 10ppm, 15ppm, 20ppm, 25ppm, 30ppm, 35ppm, 30ppm, 35ppm, 40ppm, 45ppm, 50ppm, 60ppm, 70ppm, 80ppm, 80ppm, 90ppm and 100ppm. All calculation analysis will be done using Monte Carlo MCNP6 with continuous neutron energy cross section taken from ENDF/B-VII file. Investigation of multiplication factor effect due to thermal neutron scattering crossing data S(

Item Type: Conference or Workshop Item (Paper)
Subjects: Keselamatan dan Keamanan Nuklir
Divisions: Pusat Teknologi dan Keselamatan Reaktor Nuklir
Depositing User: USER PTKRN BATAN
Date Deposited: 28 Jul 2020 06:39
Last Modified: 28 Jul 2020 06:39
URI: http://repo-nkm.batan.go.id/id/eprint/9906

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