Obtaining data on fast neutron energy spectrum by a method of differential cross-sections in recording recoil nuclei

2011 ◽  
Author(s):  
O.G. Egorov ◽  
V.V. Frunze
Keyword(s):  
Energy Spectrum ◽  
Fast Neutron ◽  
Neutron Energy ◽  
Cross Sections ◽  
Differential Cross ◽  
Neutron Energy Spectrum ◽  
Differential Cross Sections

The fast neutron energy spectrum and the thermal neutron flux distribution in the experimental channel of the VVR reactor

Atomic Energy ◽  
1962 ◽  
Vol 12 (2) ◽  
pp. 127-132 ◽  
Author(s):  
N. V. Zvonov ◽  
A. I. Mis'kevich ◽  
I. V. Rogozhkin ◽  
V. I. Tereshchenko ◽  
Zh. I. Turkov ◽  
...  
Keyword(s):  
Energy Spectrum ◽  
Neutron Flux ◽  
Thermal Neutron ◽  
Fast Neutron ◽  
Neutron Energy ◽  
Flux Distribution ◽  
Thermal Neutron Flux ◽  
Neutron Energy Spectrum ◽  
Experimental Channel ◽  
Neutron Flux Distribution

Measurement of the Fast Neutron Energy Spectrum of an $^{241}{\rm Am\!-\!Be}$ Source Using a Neutron Scatter Camera

2011 ◽  
Vol 58 (5) ◽  
pp. 2426-2430 ◽  
Author(s):  
James Brennan ◽  
Erik Brubaker ◽  
Robert Cooper ◽  
Mark Gerling ◽  
Charles Greenberg ◽  
...  
Keyword(s):  
Energy Spectrum ◽  
Fast Neutron ◽  
Neutron Energy ◽  
Neutron Energy Spectrum ◽  
Neutron Scatter ◽  
Neutron Scatter Camera

Measurement of Thermal Neutron Energy Spectrum and Fluence of the Thermal Column in Xi’an Pulse Reactor

2010 ◽  
Author(s):  
Yunhong Zhong ◽  
Xinbiao Jiang ◽  
Qingyu Yu ◽  
Wenshou Zhang
Keyword(s):  
Energy Spectrum ◽  
Thermal Neutron ◽  
Neutron Energy ◽  
Cross Sections ◽  
Neutron Fluence ◽  
Average Energy ◽  
Standard Uncertainty ◽  
Neutron Energy Spectrum ◽  
Pulse Reactor ◽  
Thermal Column

Neutron energy spectrum and fluence of the devices in a reactor are important parameters for the users. In this paper, the thermal neutron spectrum of the thermal column in Xi’an pulse reactor were measurement by the method of time of flight (TOF), and the peak of the energy spectrum is 0.0248±0.0006eV and the neutrons average energy is 0.048±0.001eV. With the spectrum results, the average cross section of 235U(n,f) and 197Au(n,γ) were calculated are respectively 92.08barn and 538.86barn. With the average cross sections, we measured the neutron fluence with the fission chamber and 197Au(n, γ) activation method. The measurements of the two methods are respectively 1.08×109n/m2s and 1.13×109n/m2s. We also calculated the standard uncertainty of the measurements: the energy spectrums’ is less than 5%, and the neutron fluence’ is less than 2.2%.

scholarly journals From Molecular Dynamics to Kinetic Rate Theory: A Simple Example of Multiscale Modeling

1998 ◽  
Vol 538 ◽  
Author(s):  
Roger E. Stoller ◽  
Lawrence R. Greenwood
Keyword(s):  
Molecular Dynamics ◽  
Energy Spectrum ◽  
Multiscale Modeling ◽  
Neutron Energy ◽  
Cross Sections ◽  
Source Term ◽  
Pressure Vessels ◽  
Dynamics Simulation ◽  
Rate Theory ◽  
Neutron Energy Spectrum

AbstractRadiation damage formation in iron has been investigated using the method of molecular dynamics simulation. The MD simulations have been used to determine primary defect production parameters for cascade energies up to 50 keV at temperatures from 100 to 900K. The energy dependence of these parameters has been used to determine appropriate neutron-energy-spectrum averaged damage production cross sections for various irradiation environments. Two applications of these effective cross sections are discussed. The first is an evaluation of neutron energy spectrum effects in commercial fission reactor pressure vessels. The second example deals with the use of these cross sections in the source term of a kinetic model used to predict void swelling and microstructural evolution. The simulation of the primary damage event by MD involves times less than 100 ps and a size scale of a few tens of nm, while the kinetic simulation encompasses several years and macroscopic sizes. This use of the MD results to develop an improved source term for rate theory modeling provides a simple example of multiscale modeling.

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