On-the-Fly Treatment of Discrete Representation Thermal Neutron Scattering Data in RMC Code

Author(s):  
Lei Zheng ◽  
Kaiwen Li ◽  
Kan Wang

Abstract A proper treatment of thermal neutron scattering data is required for the high-fidelity neutronics calculation of thermal reactors. Monte Carlo codes typically use an S(α, β) treatment to describe scattering events in the thermal region if the S(α, β) data is available for the material. The S(α,β) model stores a large majority of scattering physics and can handle thermal scattering process accurately. In neutronic-thermohydraulic coupling calculations, the temperature effect on nuclear data must be treated properly. The on-the-fly sampling method or the on-the-fly interpolation method are typically used in thermal region. In this paper, the on-the-fly interpolation method for the discrete representation S(α,β) data was introduced. The two-dimensional linear-linear interpolation was used to calculate the scattering cross sections and the secondary information for inelastic scattering, coherent elastic scattering and incoherent elastic scattering. The implemented on-the-fly capability was tested by a series of benchmarks that contain various thermal materials, including light water, beryllium and beryllium oxide. The integral kinf eigenvalues, the efficiency and the fine energy spectra of the on-the-fly treatment capacity were compared with those of the references. Results show that the on-the-fly treatment capability has high accuracy, and the computational time increases up to 20%.

2021 ◽  
Vol 247 ◽  
pp. 09012
Author(s):  
Lei Zheng ◽  
Zhiyuan Feng ◽  
Kan Wang

Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future.


2017 ◽  
Vol 146 ◽  
pp. 13009
Author(s):  
Jia Wang ◽  
Hongzhou Song ◽  
Zehua Hu ◽  
Tao Ye ◽  
Weili Sun

2013 ◽  
Vol 50 (4) ◽  
pp. 419-424 ◽  
Author(s):  
Longwei Mei ◽  
Xiangzhou Cai ◽  
Dazhen Jiang ◽  
Jingen Chen ◽  
Wei Guo ◽  
...  

2020 ◽  
Vol 239 ◽  
pp. 14008
Author(s):  
Yafen Liu ◽  
Wenjiang Li ◽  
Rui Yan ◽  
Yang Zou ◽  
Shihe Yu ◽  
...  

Thermal neutron scattering data has an important influence on the calculation and design of reactor with a thermal spectrum. However, as the only liquid fuel in the Gen-IV reactor candidates, the research on the thermal neutron scattering effect of coolant and somewhat moderator FLiBe has not been carried out sufficiently either experimentally or theoretically. The effect of FLiBe thermal neutron scattering on reactivity of TMSR-LF (thorium molten salt reactor - liquid fuel), TMSR-SF (thorium molten salt reactor - solid fuel) and MSRE (molten salt reactor experiment) were investigated and compared. Results show that the effect of FLiBe thermal neutron scattering on reactivity depends to some extent on the fuel-graphite volume ratio of core. Calculations indicate that FLiBe thermal neutron scattering of MSRE (with the hardest spectrum) has the minimum effect of 41 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-SF (with the softest spectrum) has the maximum effect of -94 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-LF has an effect of -61 pcm on reactivity at 900 K.


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