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Monte Carlo Simulation in Reactor Physics
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Monte Carlo Simulation in Reactor Physics

A special issue of Journal of Nuclear Engineering (ISSN 2673-4362).

Deadline for manuscript submissions: closed (31 July 2024) | Viewed by 12261

Special Issue Editor

Dr. Shichang Liu

E-Mail Website
Guest Editor
School of Nuclear Science and Engineering, North China Electric Power University, Beijing 102206, China
Interests: advanced nuclear systems; monte carlo; particle transport; fusion neutronics; radiation protection; multi-physical coupling
Special Issues, Collections and Topics in MDPI journals

Special Issue Information

Dear Colleagues,

With the increasing demand for high-fidelity neutronics analysis and the development of computer technology, the Monte Carlo method becomes increasingly important, especially in the critical analysis of initial core and shielding calculations. This is due to its advantages such as flexibility in geometry treatment, the ability to use continuous-energy pointwise cross sections, the ease of parallelization, and the high fidelity of simulations. In the last year, many new-generation Monte Carlo codes were developed, including MCNP6, OpenMC, MC21, SHIFT, TRIPOLI, Geant4, Spernt, MCCARD, MCS, RMC, SuperMC, JMCT, etc. These codes are aimed at achieving full core calculations and analyses with high fidelity and efficiency by means of advanced methodologies and algorithms as well as high-performance computing techniques.

This Special Issue of JNE will publish some of the most recent advances in the field of algorithms and technological developments in new-generation Monte Carlo codes. The landscape of the subject is very wide, and therefore it was necessary to limit the selection of topics. This selection is based mainly on the competences of the Guest Editor. This volume will include some of the most recent as well as state-of-the-art methodologies, simulations, and applications on fission neutronics, fusion neutronics, particle transport, multi-physical coupling, shielding calculations, and high-performance computing.

Dr. Shichang Liu
Guest Editor

Manuscript Submission Information

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Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Journal of Nuclear Engineering is an international peer-reviewed open access quarterly journal published by MDPI.

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Keywords

  • Monte Carlo method
  • particle transport
  • reactor physics
  • fusion neutronics
  • fission neutronics
  • multi-physical coupling
  • shielding calculations
  • high-performance computing

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Published Papers (6 papers)

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Research

25 pages, 8740 KiB  
Article
Open-Source Optimization of Hybrid Monte Carlo Methods for Fast Response Modeling of NaI (Tl) and HPGe Gamma Detectors
by Matthew Niichel and Stylianos Chatzidakis
J. Nucl. Eng. 2024, 5(3), 274-298; https://doi.org/10.3390/jne5030019 - 5 Aug 2024
Viewed by 776
Abstract
Modeling the response of gamma detectors has long been a challenge within the nuclear community. Significant research has been conducted to digitally replicate instruments that can cost over USD 100,000 and are difficult to operate outside of a laboratory setting. The large cost [...] Read more.
Modeling the response of gamma detectors has long been a challenge within the nuclear community. Significant research has been conducted to digitally replicate instruments that can cost over USD 100,000 and are difficult to operate outside of a laboratory setting. The large cost and availability prevent some from making use of such equipment. Subsequently, there have been multiple attempts to create cost-effective codes that replicate the response of sodium-iodide and high-purity germanium detectors for data derivation related to gamma-ray interaction with matter. While robust programs do exist, they are often subject to export controls and/or they are not intuitive to use. Through the use of hybrid Monte Carlo methods, MATLAB can be used to produce a fast first-order response of various gamma-ray detectors. The combination of a graphical user interface with a numerical-based script allows for open-source and intuitive code. When benchmarked with experimental data from Co-60, Cs-137, and Na-22, the code can numerically calculate a response comparable to experimental and industry-standard response codes. Evidence supports both savings in computational requirements and the inclusion of an intuitive user experience that does not heavily compromise data when compared to other standard codes, such as MCNP and GADRAS, or experimental results. When the application is installed on a Dell Intel i7 computer with 16 cores, the average time to simulate the benchmarked isotopes is 0.26 s. Installation on an HP Intel i7 four-core machine runs the same isotopes in 1.63 s. The results indicate that simple gamma detectors can be modeled in an open-source format. The anticipation for the MATLAB application is to be a tool that can be easily accessible and provide datasets for use in an academic setting requiring gamma-ray detectors. Ultimately, this article provides evidence that hybrid Monte Carlo codes in an open-source format can benefit the nuclear community in both computational time and up-front cost for access. Full article
(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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12 pages, 4480 KiB  
Article
Research on the Influence of Negative KERMA Factors on the Power Distribution of a Lead-Cooled Fast Reactor
by Guanqun Jia, Xubo Ma, Teng Zhang and Kui Hu
J. Nucl. Eng. 2024, 5(1), 1-12; https://doi.org/10.3390/jne5010001 - 21 Dec 2023
Viewed by 1077
Abstract
The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and [...] Read more.
The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark. Full article
(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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20 pages, 9355 KiB  
Article
Estimation of Continuous Distribution of Iterated Fission Probability Using an Artificial Neural Network with Monte Carlo-Based Training Data
by Delgersaikhan Tuya and Yasunobu Nagaya
J. Nucl. Eng. 2023, 4(4), 691-710; https://doi.org/10.3390/jne4040043 - 6 Nov 2023
Viewed by 1649
Abstract
The Monte Carlo neutron transport method is used to accurately estimate various quantities, such as k-eigenvalue and integral neutron flux. However, in the case of estimating a distribution of a desired quantity, the Monte Carlo method does not typically provide continuous distribution. Recently, [...] Read more.
The Monte Carlo neutron transport method is used to accurately estimate various quantities, such as k-eigenvalue and integral neutron flux. However, in the case of estimating a distribution of a desired quantity, the Monte Carlo method does not typically provide continuous distribution. Recently, the functional expansion tally (FET) and kernel density estimation (KDE) methods have been developed to provide a continuous distribution of a Monte Carlo tally. In this paper, we propose a method to estimate a continuous distribution of a quantity in all phase-space variables using a fully connected feedforward artificial neural network (ANN) model with Monte Carlo-based training data. As a proof of concept, a continuous distribution of iterated fission probability (IFP) was estimated by ANN models in two distinct fissile systems. The ANN models were trained on the training data created using the Monte Carlo IFP method. The estimated IFP distributions by the ANN models were compared with the Monte Carlo-based data that include the training data. Additionally, the IFP distributions by the ANN models were also compared with the adjoint angular neutron flux distributions obtained with the deterministic neutron transport code PARTISN. The comparisons showed varying degrees of agreement or discrepancy; however, it was observed that the ANN models learned the general trend of the IFP distributions from the Monte Carlo-based training data. Full article
(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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8 pages, 1392 KiB  
Communication
Preliminary Study on the Thermal Neutron Scattering Cross-Section for HinH2O in Small Modular Reactors
by Jun Wu and Yixue Chen
J. Nucl. Eng. 2023, 4(2), 309-316; https://doi.org/10.3390/jne4020023 - 4 Apr 2023
Viewed by 1697
Abstract
Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for [...] Read more.
Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation. Full article
(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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11 pages, 3394 KiB  
Article
Investigating Sustainability Index, 99Mo Output and 239Pu Levels in UO2 Targets by Substituting 238U with Ce
by Robert Raposio, Anatoly Rosenfeld, Juniper Bedwell-Wilson and Gordon Thorogood
J. Nucl. Eng. 2022, 3(4), 295-305; https://doi.org/10.3390/jne3040017 - 26 Oct 2022
Viewed by 2070
Abstract
A new target material combination was modelled to replace the existing uranium-aluminium design used for 99Mo manufacture to increase the sustainability of the production process. Previous efforts to develop a more sustainable uranium target for 99Mo production, resulted in the levels [...] Read more.
A new target material combination was modelled to replace the existing uranium-aluminium design used for 99Mo manufacture to increase the sustainability of the production process. Previous efforts to develop a more sustainable uranium target for 99Mo production, resulted in the levels of 239Pu in the target after irradiation being elevated due to the increase in 238U present. MCNP6.2 was used to model 4 different cylindrical targets based on 4–7 days irradiation to further understand this effect. To reduce the resultant 239Pu levels, ratios of 0–99% of Ce were used as a replacement for 238U. The results show that the addition of 140Ce and the removal of 238U reduced the 239Pu levels in the target significantly thus increasing the sustainability of the target and giving a slight increase to the 99Mo output of the targets. Full article
(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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11 pages, 1280 KiB  
Article
Efficiency Studies of Fast Neutron Tracking Using MCNP
by Pinghan Chu, Michael R. James and Zhehui Wang
J. Nucl. Eng. 2022, 3(2), 117-127; https://doi.org/10.3390/jne3020007 - 30 Apr 2022
Cited by 2 | Viewed by 2785
Abstract
Fast neutron identification and spectroscopy is of great interest to nuclear physics experiments. Using the neutron elastic scattering, the fast neutron momentum can be measured. Wang and Morris introduced the theoretical concept that the initial fast neutron momentum can be derived from up [...] Read more.
Fast neutron identification and spectroscopy is of great interest to nuclear physics experiments. Using the neutron elastic scattering, the fast neutron momentum can be measured. Wang and Morris introduced the theoretical concept that the initial fast neutron momentum can be derived from up to three consecutive elastic collisions between the neutron and the target, including the information of two consecutive recoil ion tracks and the vertex position of the third collision or two consecutive elastic collisions with the timing information. Here, we also include the additional possibility of measuring the deposited energies from the recoil ions. In this paper, we simulate the neutron elastic scattering using the Monte Carlo N-Particle Transport Code (MCNP) and study the corresponding neutron detection and tracking efficiency. The corresponding efficiency and the scattering distances are simulated with different target materials, especially natural silicon (92.23%28Si, 4.67%29Si, and 3.1%30Si) and helium-4 (4He). The timing of collision and the recoil ion energy are also investigated, which are important characters for the detector design. We also calculate the ion traveling range for different energies using the software, “The Stopping and Range of Ions in Matter (SRIM)”, showing that the ion track can be most conveniently observed in 4He unless sub-micron spatial resolution can be obtained in silicon. Full article
(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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