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Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents

Abstract

It presents the experimental measurements of cross section of fissile nuclei for slow neutrons to improve the understanding of some heavy nuclei of great importance in the study of nuclear reactors. The different experiments are divided in three categories. In the first part, it studied the variation with energy of the cross sections of natural uranium, {sup 233}U, {sup 235}U and {sup 239}Pu. Two measurement techniques are used: the time-of-flight spectrometer and the crystal spectrometer. In a second part, the fission cross sections of {sup 233}U and {sup 239}Pu for thermal neutrons are compared using a neutron flux from EL-2 going through a double fission chamber. The matter quantity contained in each source is measured by counting the {alpha} activity with a solid angle counter. Finally, the average cross section of {sup 236}U for a spectra of neutrons from the reactor is measured by studying the {beta} activity of {sup 237}U formed by the reaction {sup 236}U (n, {gamma}) {sup 237}U in a sample of {sup 236}U irradiated in the Saclay reactor (EL-2). (M.P.)
Authors:
Auclair, J M; Hubert, P; Joly, R; Vendryes, G; Jacrot, B; Netter, F; [1]  Galula, M [2] 
  1. Commissariat a l'Energie Atomique, Saclay (France). Centre d'Etudes Nucleaires
  2. Centre National de la Recherche Scientifique (CNRS), 91 - Gif-sur-Yvette (France)
Publication Date:
Jul 01, 1955
Product Type:
Technical Report
Report Number:
CEA-R-446
Resource Relation:
Conference: Geneva conference, Conference de Geneve, Geneva (Switzerland), Aug 1955; Other Information: 7 refs
Subject:
73 NUCLEAR PHYSICS AND RADIATION PHYSICS; ALPHA DECAY; BF3 COUNTERS; EPITHERMAL NEUTRONS; EV RANGE; FISSION CHAMBERS; FISSION PRODUCTS; NATURAL URANIUM; NEUTRON BEAMS; NEUTRON DIFFRACTION; NEUTRON FLUX; NEUTRON REACTIONS; PLUTONIUM 239; SLOW NEUTRONS; THERMAL NEUTRONS; TIME-OF-FLIGHT SPECTROMETERS; TOTAL CROSS SECTIONS; URANIUM 233; URANIUM 235; URANIUM 236
OSTI ID:
20868348
Research Organizations:
CEA Saclay, 91 - Gif-sur-Yvette (France)
Country of Origin:
France
Language:
French
Other Identifying Numbers:
TRN: FR07R0446036101
Availability:
Available from INIS in electronic form
Submitting Site:
FRN
Size:
35 pages
Announcement Date:
May 24, 2007

Citation Formats

Auclair, J M, Hubert, P, Joly, R, Vendryes, G, Jacrot, B, Netter, F, and Galula, M. Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents. France: N. p., 1955. Web.
Auclair, J M, Hubert, P, Joly, R, Vendryes, G, Jacrot, B, Netter, F, & Galula, M. Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents. France.
Auclair, J M, Hubert, P, Joly, R, Vendryes, G, Jacrot, B, Netter, F, and Galula, M. 1955. "Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents." France.
@misc{etde_20868348,
title = {Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents}
author = {Auclair, J M, Hubert, P, Joly, R, Vendryes, G, Jacrot, B, Netter, F, and Galula, M}
abstractNote = {It presents the experimental measurements of cross section of fissile nuclei for slow neutrons to improve the understanding of some heavy nuclei of great importance in the study of nuclear reactors. The different experiments are divided in three categories. In the first part, it studied the variation with energy of the cross sections of natural uranium, {sup 233}U, {sup 235}U and {sup 239}Pu. Two measurement techniques are used: the time-of-flight spectrometer and the crystal spectrometer. In a second part, the fission cross sections of {sup 233}U and {sup 239}Pu for thermal neutrons are compared using a neutron flux from EL-2 going through a double fission chamber. The matter quantity contained in each source is measured by counting the {alpha} activity with a solid angle counter. Finally, the average cross section of {sup 236}U for a spectra of neutrons from the reactor is measured by studying the {beta} activity of {sup 237}U formed by the reaction {sup 236}U (n, {gamma}) {sup 237}U in a sample of {sup 236}U irradiated in the Saclay reactor (EL-2). (M.P.)}
place = {France}
year = {1955}
month = {Jul}
}