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%.

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

Thermal Neutron Energy Spectrum of JRR-3

1969 ◽  
Vol 6 (6) ◽  
pp. 341-343
Author(s):  
Yuji ISHIGURO ◽  
Shinji AYAO ◽  
Mitsuho J. HIRATA
Keyword(s):  
Energy Spectrum ◽  
Thermal Neutron ◽  
Neutron Energy ◽  
Neutron Energy Spectrum

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.

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

1998 ◽  
Vol 540 ◽  
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.

Space to Air High-Altitude Region Adjoint Neutron Transport

2021 ◽  
pp. 154851292110316
Author(s):  
Zachary W LaMere ◽  
Darren E Holland ◽  
Whitman T Dailey ◽  
John W McClory
Keyword(s):  
Energy Spectrum ◽  
High Altitude ◽  
Neutron Energy ◽  
Neutron Transport ◽  
Energy Sources ◽  
Source Spectrum ◽  
Detector Response ◽  
Neutron Energy Spectrum ◽  
True Source ◽  
Adjoint Transport

Neutrons from an atmospheric nuclear explosion can be detected by sensors in orbit. Current tools for characterizing the neutron energy spectrum assume a known source and use forward transport to recreate the detector response. In realistic scenarios the true source is unknown, making this an inefficient, iterative approach. In contrast, the adjoint approach directly solves for the source spectrum, enabling source reconstruction. The time–energy fluence at the satellite and adjoint transport equation allow a Monte Carlo method to characterize the neutron source’s energy spectrum directly in a new model: the Space to High-Altitude Region Adjoint (SAHARA) model. A new adjoint source event estimator was developed in SAHARA to find feasible solutions to the neutron transport problem given the constraints of the adjoint environment. This work explores SAHARA’s development and performance for mono-energetic and continuous neutron energy sources. In general, the identified spectra were shifted towards energies approximately 5% lower than the true source spectra, but SAHARA was able to capture the correct spectral shapes. Continuous energy sources, including real-world sources Fat Man and Little Boy, resulted in identifiable spectra that could have been produced by the same distribution as the true sources as demonstrated by two-dimensional (2D) Kolmogorov–Smirnov tests.

Sign in / Sign up

Export Citation Format

Share Document