| Popis: |
In the complex realm of nuclear engineering, understanding the kinetics behavior of neutrons in reactors is a task of considerable importance for safety. The effective delayed neutron fraction (βeff), neutron prompt generation time (Λ) and other neutron kinetics parameters are pivotal for dynamics analysis in nuclear reactor. Because of the limited realistic conditions, these parameters have traditionally been challenging to measure experimentally, compelling researchers to rely on numerical computations for their determination, especially for pebble-bed high-temperature gas-cooled reactors (HTGRs). Yet conventional numerical techniques necessitate direct core condition data and often presuppose simplistic treatments of delayed neutron energies, resulting in computational inefficiencies and imprecise outcomes. A central difficulty lies with the process involving VSOP and CITATION codes, used in HTGR calculations of neutron kinetics parameters, which requires the formulation of input cards based on VSOP output for any core state—This procedure can confound researchers. Furthermore, the limited data library (ENDF/B-Ⅵ.8) available to VSOP restricts the accuracy attainable through these methods. In this paper, in an effort to refine this process, a multi-group cross section library, XPZLIB, containing delayed neutron data was generated. This library leverages the comprehensive nuclear data from the ENDF/B-Ⅷ.0 library, processed using the enhanced NJOY2016 code equipped with a developed XPZR module and an automatic processing system, PyNjoy2022. The HTGR physics lattice code XPZ was then used to homogenize the delayed neutron data, ensuring that the necessary energy-related information was retained in preparation for subsequent calculations of energy-dependent neutron kinetic parameters. A unique homogenization technique was also developed to handle the delayed neutron precursor decay constants (λ), which unlike other parameters, are time-dependent. These calculations of kinetics parameters for HTGRs are conducted via the importance function weighting method facilitated by the core physical analysis code PANGU. This method utilizes the adjoint neutron flux as the weighting function, which is determined from the adjoint neutron transport equation. The approach produced a collection of single-speed neutron kinetic parameters as well as a 6-group energy-dependent effective delayed neutron fraction. The accuracy and validity of this methodology were confirmed by benchmarking the results against Monte Carlo solutions using iterated fission probability method. The findings not only confirm that comprehensive numerical calculations are feasible for determining neutron kinetics parameters in HTGR environments but also provide a robust foundation for subsequent dynamics analyses. By contrasting the new method with pre-existing pre-fitting methods, the superiority of this advanced technique became evident, signifying a significant stride forward in the field of reactor dynamics and safety analysis. |
