The total neutron cross section for natural carbon in the energy range 2 to 148 keV

Abstract. An experimental investigations of the total neutron cross section for natural carbon were done at Kyiv Research Reactor using neutron filtered beams with energies 2, 3.5, 12, 24, 55, 59, 133 and 148 keV. The intense neutron beams formed by composite neutron filters at reactor horizontal channels had the fluxes of about 10 6 -10 7 neutron/cm 2. s at the fixed neutron energies and this enabled to measure the neutron cross sections with accuracy better than 3%. Transmission method was used in these measurements. The results are presented together with the analysis of the known previous experimental data and the evaluated nuclear data from ENDF libraries

Neutron Scattering Cross Sections for Natural Carbon in the Energy Range 2-133 keV

Natural carbon is well known as reactor structure material and at the same time as one of the most important neutron scattering standards, especially at energies less than 2 MeV, where the neutron total and neutron scattering cross sections are essentially identical. The best neutron total cross section experimental data for natural carbon in the range 1-500 keV have uncertainties of 1-4%. However, the difference between these data and those based on R-matrix analysis and used in the ENDF libraries is evident, especially in the energy range 1-60 keV. Experimental data for total scattering neutron cross sections for this element in the energy range 1-200 keV are scanty. The use of the technique of neutron filtered beams developed at the Kyiv Research Reactor makes it possible to reduce the uncertainty of the experimental data and to measure the neutron scattering cross sections on natural carbon in the energy range 2-149 keV with accuracies of 3-6%. Investigations of the neutron scattering cross section on carbon were carried out using 5 filters with energies 2, 3.5, 24, 54 and 133 keV. The neutron scattering cross sections were measured using a detector system covering nearly 2{pi}. The detector consisting of {sup 3}He counters (58 units), was located just above the carbon samples. The {sup 3}He counters (CHM-37, 7 atm, diameter =18 mm, L=50 cm) are placed in five layers (12 or 11 in each layer). To determine the neutron scattering cross section on carbon the relative method of measurement was used. The isotope {sup 208}Pb was used as the standard. The normalization factor, which is a function of detector efficiency, thickness of the carbon samples, thickness of the {sup 208}Pb sample, geometry, etc., for each sample and for each filter energy has been obtained through Monte Carlo calculations by means of the MCNP4C code. The results of measurements of the neutron scattering cross sections at reactor neutron filtered beams with energies in the range 2-133 keV on carbon samples together with the known experimental data from database EXFOR/CSISRS and ENDF libraries are presented

Investigation of the scattering cross sections of neutrons on carbon nuclei at the reactor filtered beams

Natural carbon is well known as reactor structure material and at the same time as one of the most important neutron scattering standards, especially at energies less than 2 MeV, where the neutron total and neutron scattering cross sections are essentially identical. The best neutron total cross section experimental data for natural carbon in the range 1 - 500 keV has uncertainties of 1 - 4 %. However, the difference between these data and those based on R-matrix analysis and used in the ENDF libraries is evident; especially in the energy range 1 - 60 keV. Experimental data for total scattering neutron cross sections for this element in the energy range 1 - 200 keV are scanty. The use of the technique of neutron filtered beams developed at the Kyiv Research Reactor makes it possible to reduce the uncertainty of the experimental data and to measure the neutron scattering cross sections on natural carbon in the energy range 2 - 149 keV with accuracies of 3 - 6 %. Investigations of the neutron scattering cross section on carbon were carried out using 5 filters with energies 2, 3.5, 24, 54 and 133 keV. The neutron scattering cross sections were measured using a detector system covering nearly 2π. The detector consisting of 3 He counters (58 units), was located just above the carbon samples. The 3 He counters (CHM-37, 7 atm, diameter = 18 mm, L = 50 cm) are placed in five layers (12 or 11 in each layer). To determine the neutron scattering cross section on carbon the relative method of measurement was used. The isotope 208Pb was used as the standard. The normalization factor, which is a function of detector efficiency, thickness of the carbon samples, thickness of the 208Pb sample, geometry, etc., for each sample and for each filter energy has been obtained through Monte Carlo calculations by means of the MCNP4C code. The results of measurements of the neutron scattering cross sections at reactor neutron filtered beams with energies in the range 2 - 133 keV on carbon samples together with the known experimental data from database EXFOR/CSISRS and ENDF libraries are presented

Towards a More Complete and Accurate Experimental Nuclear Reaction Data Library (EXFOR): International Collaboration Between Nuclear Reaction Data Centres (NRDC)

The International Network of Nuclear Reaction Data Centres (NRDC) coordinated by the IAEA Nuclear Data Section (NDS) is successfully collaborating in the maintenance and development of the EXFOR library. As the scope of published data expands (e.g., to higher energy, to heavier projectile) to meet the needs from the frontier of sciences and applications, it becomes nowadays a hard and challenging task to maintain both completeness and accuracy of the whole EXFOR library. The paper describes evolution of the library with highlights on recent developments.Comment: 4 pages, 2 figure

Modeling of the new neutron filter with 5.6 keV energy

New filtered neutron beam with the average energy of 5.6 keV was simulated using the latest versions of the evaluated nuclear data libraries ENDF/B-VII.1 and CENDL-3.1. The main components of this new filter are: 60Ni, manganese, sulfur, and cerium oxide. Expected filter characteristics are: the average energy is 5.62 keV; neutron line width at half maximum is 1.73 keV; the purity of the main line is 84 %; the neutron flux is 2⋅105 n/(s⋅cm2). Experimental testing of the parameters of this new filter will be done during the next campaign of the reactor WWR-M

Present Status of Neutron-, Photo-induced and Spontaneous Fission Yields Experimental Data

Nuclear reaction data collection, evaluation and dissemination have been pioneered at the Brookhaven National Laboratory since the early 50s. These activities gained popularity worldwide, and around 1970 the experimental nuclear reaction data interchange or exchange format (EXFOR) was established. The original EXFOR compilation scope consisted only of neutron reactions and spontaneous fission data, while many other nuclear data sets were ignored. Due to the high cost of new experiments, it is very important to find and recover the previously disregarded data using scientific publications, data evaluations and nuclear databases comparisons. Fission yields play a very important role in applied and fundamental physics, and such data are essential in many applications. The comparative analysis of Nuclear Science References (NSR) and Experimental Nuclear Reaction (EXFOR) databases shows a large number of unaccounted experiments and provides a guide for the recovery of fission cross sections, yields and covariance data sets. The dedicated fission yields data compilation effort is currently underway in the Nuclear Reaction Data Centers (NRDC) network, and includes identification, compilation, storage and Web dissemination of the recovered data sets

Present Status of Neutron-, Photo-induced and Spontaneous Fission Yields Experimental Data

Nuclear reaction data collection, evaluation and dissemination have been pioneered at the Brookhaven National Laboratory since the early 50s. These activities gained popularity worldwide, and around 1970 the experimental nuclear reaction data interchange or exchange format (EXFOR) was established. The original EXFOR compilation scope consisted only of neutron reactions and spontaneous fission data, while many other nuclear data sets were ignored. Due to the high cost of new experiments, it is very important to find and recover the previously disregarded data using scientific publications, data evaluations and nuclear databases comparisons. Fission yields play a very important role in applied and fundamental physics, and such data are essential in many applications. The comparative analysis of Nuclear Science References (NSR) and Experimental Nuclear Reaction (EXFOR) databases shows a large number of unaccounted experiments and provides a guide for the recovery of fission cross sections, yields and covariance data sets. The dedicated fission yields data compilation effort is currently underway in the Nuclear Reaction Data Centers (NRDC) network, and includes identification, compilation, storage and Web dissemination of the recovered data sets

Methods of experimental settlement of contradicting data in evaluated nuclear data libraries

The latest versions of the evaluated nuclear data libraries (ENDLs) have contradictions concerning data about neutron cross sections. To resolve this contradiction we propose the method of experimental verification. This method is based on using of the filtered neutron beams and following measurement of appropriate samples. The basic idea of the method is to modify the suited filtered neutron beam so that the differences between the neutron cross sections in accordance with different ENDLs become measurable. Demonstration of the method is given by the example of cerium, which according to the latest versions of four ENDLs has significantly different total neutron cross section