Energy response based neutron spectrum unfolding method
By employing an energy response-based neutron spectrum analysis method, which utilizes response matrix and iterative matrix operations, the problems of low efficiency and low resolution in existing neutron spectrum analysis technologies are solved, achieving efficient and high-precision neutron spectrum analysis.
CN119132475BActive Publication Date: 2026-07-21SHANGHAI JIAOTONG UNIV
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Patent Information
- Authority / Receiving Office
- CN · China
- Patent Type
- Patents(China)
- Current Assignee / Owner
- SHANGHAI JIAOTONG UNIV
- Filing Date
- 2024-09-12
- Publication Date
- 2026-07-21
AI Technical Summary
Technical Problem
Existing methods for analyzing neutron spectra are computationally inefficient and have low spectral resolution, making it impossible to obtain high-precision neutron spectra directly through simulation.
Method used
The neutron spectrum analysis method based on energy response generates a response matrix by analyzing a nuclear database. It then combines the initial source term and the iteration of the response matrix to quickly analyze the neutron spectrum, using Monte Carlo methods and matrix operations for high-resolution analysis.
Benefits of technology
It achieves accurate analysis of high-resolution neutron energy spectra, significantly improves computational efficiency, reduces approximation and statistical errors, and ensures high accuracy of results.
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Figure CN119132475B_ABST
Abstract
The application discloses a neutron energy spectrum analysis method based on energy response, which comprises the following steps: generating an energy response matrix of a target material according to nuclide proportioning and weighting; modeling a fuel rod area and taking an average collision probability matrix as an interpolation factor to interpolate incident neutron energy distribution when the cross section of the fuel rod model is 0 and infinite; obtaining the incident neutron energy distribution; performing matrix multiplication operation on the incident neutron energy distribution, the energy response matrix of the target material and the average collision probability matrix to obtain the neutron energy distribution after one collision and update the incident neutron energy distribution as one collision simulation; and dividing each element in the incident neutron energy distribution by a corresponding energy group total cross section to obtain the neutron energy spectrum after multiple collision simulations until the incident neutron energy distribution is stable.
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