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State-of-Art in Nuclear Reactor Physics
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State-of-Art in Nuclear Reactor Physics

A special issue of Energies (ISSN 1996-1073). This special issue belongs to the section "B4: Nuclear Energy".

Deadline for manuscript submissions: closed (10 August 2022) | Viewed by 19477

Special Issue Editor

Dr. Dan Kotlyar

E-Mail Website
Guest Editor
Georgia Institute of Technology, G. W. Woodruff School of Mechanical Engineering, Nuclear and Rediological Engineering Program, North Ave NW, Atlanta, GA 30332, USA
Interests: reactor phycics; fuel cycles; reactor engineering; multiphysics
Special Issues, Collections and Topics in MDPI journals

Special Issue Information

Dear Colleagues,

There are many technical challenges relating to the licensing of advanced concepts and promoting new fuel and clad technologies. In addition, the regulatory requirements for safety criteria are constantly evolving. Utilities are also facing various engineering challenges that include spent fuel management, irradiation-induced reactor vessel damage, and many others. Advanced methods and integrated/coupled tools are necessary to understand the operational limits and assess fuel performance, in terms of burnup and thermal hydraulic reliability, and the associated safety margins. High-fidelity computational codes and multiphysics platforms (e.g., Nuclear Energy Advanced Modeling and Simulation (NEAMS)) are of the utmost importance in capturing important physical effects. An example of the latter is hydrogen permeation or loss from hydride moderators, such as zirconium hydride or yttrium hydride, which represents a critical design challenge for hydride-moderated nuclear reactors. This Special Issue of Energies on “State-of-Art in Nuclear Reactor Physics” focuses on new methodologies, techniques, and computational frameworks that are directly applicable to solve various design and licensing problems of modern nuclear reactor fission systems.

Relevant topics include, but are not limited to, the following research topics:

  • State-of-the-art spatial kinetics methods;
  • Modern nodal diffusion codes;
  • The advancement of 2D and 3D deterministic transport methods;
  • The use of Monte Carlo methods for full-core solutions;
  • Time-dependent Monte Carlo methods;
  • The development and/or application of multiphysics frameworks.

Manuscripts addressing any combination of these topics where more than a single aspect of physics is coupled with any of the above options (e.g., coupled Monte Carlo and thermal-hydraulics or coupled nodal diffusion with Monte Carlo) are also welcome.

Dr. Dan Kotlyar
Guest Editor

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

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. Energies is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2600 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • Neutron diffusion
  • Deterministic transport
  • Monte Carlo
  • Multiphysics
  • Kinetics and dynamics.

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Published Papers (6 papers)

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Research

25 pages, 5181 KiB  
Article
Full-Core Coupled Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Analysis of Low-Enriched Uranium Nuclear Thermal Propulsion Reactors
by Matt Krecicki and Dan Kotlyar
Energies 2022, 15(19), 7007; https://doi.org/10.3390/en15197007 - 24 Sep 2022
Cited by 14 | Viewed by 3173
Abstract
Nuclear thermal propulsion is an enabling technology for future space missions, such as crew-operated Mars missions. Nuclear thermal propulsion technology provides a performance benefit over chemical propulsion systems by operating with light propellants (e.g., hydrogen) at elevated engine chamber conditions. Therefore, nuclear thermal [...] Read more.
Nuclear thermal propulsion is an enabling technology for future space missions, such as crew-operated Mars missions. Nuclear thermal propulsion technology provides a performance benefit over chemical propulsion systems by operating with light propellants (e.g., hydrogen) at elevated engine chamber conditions. Therefore, nuclear thermal propulsion reactor cores exhibit high propellant velocities and elevated propellant and fuel temperatures, subsequently leading to relatively high thermal stresses and geometrical deformation. This paper details the numerical approach to solve the thermo-elastic equations, which was implemented into the recently developed ntpThermo code. In addition, this paper demonstrates the extension of the Basilisk multiphysics framework to perform full-core coupled neutronic, thermal-hydraulic, and thermo-mechanical analysis of nuclear thermal propulsion reactors. The analyses demonstrate and quantify thermo-mechanical feedback, which for the investigated cases, acted to reduce maximum fuel temperatures and pressure drop across the fuel element channels. Thermo-mechanical feedback had a significant impact on the mass flow distribution within the reactor core and, thus, a substantial impact on solid-material temperatures and stresses, but only a minor impact on reactivity and local power distributions. Sensitivity studies revealed that the friction factor correlation applied to perform the analysis has a significant impact on the pressure drop across the fuel element channels. The most important observation of this research is the importance of incorporating the thermo-mechanical feedback within an integrated multiphysics solution sequence to enable the consistent design of future nuclear thermal propulsion systems. Full article
(This article belongs to the Special Issue State-of-Art in Nuclear Reactor Physics)
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27 pages, 60316 KiB  
Article
Development of Explainable Data-Driven Turbulence Models with Application to Liquid Fuel Nuclear Reactors
by Mauricio E. Tano and Pablo Rubiolo
Energies 2022, 15(19), 6861; https://doi.org/10.3390/en15196861 - 20 Sep 2022
Cited by 4 | Viewed by 1991
Abstract
Liquid fuel nuclear reactors offer innovative possibilities in terms of nuclear reactor designs and passive safety systems. Molten Salts Reactors (MSRs) with a fast spectrum are a particular type of these reactors using liquid fuel. MSFRs often involve large open cavities in their [...] Read more.
Liquid fuel nuclear reactors offer innovative possibilities in terms of nuclear reactor designs and passive safety systems. Molten Salts Reactors (MSRs) with a fast spectrum are a particular type of these reactors using liquid fuel. MSFRs often involve large open cavities in their core in which the liquid fuel circulates at a high speed to transport the heat generated by the nuclear reactions into the heat exchangers. This high-speed flow yields a turbulent field with large Reynolds numbers in the reactor core. Since the nuclear power, the neutron precursor’s transport and the thermal exchanges are strongly coupled in the MSFR’s core cavity, having accurate turbulent models for the liquid fuel flow is necessary to avoid introducing significant errors in the numerical simulations of these reactors. Nonetheless, high-accuracy simulations of the turbulent flow field in the reactor cavity of these reactors are usually prohibitively expensive in terms of computational resources, especially when performing multiphysics numerical calculations. Therefore, in this work, we propose a novel method using a modified genetic algorithm to optimize the calculation of the Reynolds Shear Stress Tensor (RST) used for turbulence modeling. The proposed optimization methodology is particularly suitable for advanced liquid fuel reactors such as the MSFRs since it allows the development of high-accuracy but still low-computational-cost turbulence models for the liquid fuel. We demonstrate the applicability of this approach by developing high accuracy Reynolds-Averaged Navier–Stokes (RANS) models (averaged flow error less than 5%) for a low and a large aspect ratio in a Backward-Facing Step (BFS) section particularly challenging for RANS models. The newly developed turbulence models better capture the flow field after the boundary layer tipping, over the extent of the recirculation bubble, and near the boundary layer reattachment region in both BFS configurations. The main reason for these improvements is that the developed models better capture the flow field turbulent anisotropy in the bulk region of the BFS. Then, we illustrate the interest in using this turbulence modeling approach for the case of an MSFR by quantifying the impact of the turbulence modeling on the reactor key parameters. Full article
(This article belongs to the Special Issue State-of-Art in Nuclear Reactor Physics)
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28 pages, 9130 KiB  
Article
Multiphysics Simulation of the NASA SIRIUS-CAL Fuel Experiment in the Transient Test Reactor Using Griffin
by Tian Jing, Sebastian Schunert, Vincent M. Labouré, Mark D. DeHart, Ching-Sheng Lin and Javier Ortensi
Energies 2022, 15(17), 6181; https://doi.org/10.3390/en15176181 - 25 Aug 2022
Cited by 10 | Viewed by 2439
Abstract
After approximately 50 years, NASA is restarting efforts to develop nuclear thermal propulsion (NTP) for interplanetary missions. Building upon nuclear engine tests performed from the late 1950s to the early 1970s, the present research and testing focuses on advanced materials and fabrication methods. [...] Read more.
After approximately 50 years, NASA is restarting efforts to develop nuclear thermal propulsion (NTP) for interplanetary missions. Building upon nuclear engine tests performed from the late 1950s to the early 1970s, the present research and testing focuses on advanced materials and fabrication methods. A number of transient tests have been performed to evaluate materials performance under high-temperature, high-flux conditions, with several more experiments in the pipeline for future testing. The measured data obtained from those tests are being used to validate the Griffin reactor multiphysics code for this particular type of application. Griffin was developed at Idaho National Laboratory (INL) using the MOOSE framework. This article describes the simulation results of the SIRIUS-CAL calibration experiment in the Transient Reactor Test Facility (TREAT). SIRIUS-CAL was the first transient test conducted on NASA fuels, and although the test was performed with a relatively low core peak power, the test specimen survived a temperature exceeding 900 K. Griffin simulations of the experiment successfully matched the reactor’s power transient after calibrating the initial control rod position to match the initial reactor period. The thermal-hydraulics model largely matches the time-dependent response of a thermocouple located within the experiment specimen to within the uncertainty estimate. However, the uncertainty range is significant and must be reduced in the future. Full article
(This article belongs to the Special Issue State-of-Art in Nuclear Reactor Physics)
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11 pages, 2296 KiB  
Article
Investigation and Validation of Unstructured Mesh Methodologies for Modeling Experimental Reactors
by Tom Mager, Carlo Fiorina, Mathieu Hursin and Andreas Pautz
Energies 2022, 15(4), 1512; https://doi.org/10.3390/en15041512 - 18 Feb 2022
Cited by 1 | Viewed by 1945
Abstract
This paper summarizes a methodology developed at École Polytechnique Fédérale de Lausanne for the neutronic modeling of the CROCUS experimental reactor and proposes solutions to the challenges one may face while modeling a research reactor with a complex geometry. Indeed, the double-lattice configuration [...] Read more.
This paper summarizes a methodology developed at École Polytechnique Fédérale de Lausanne for the neutronic modeling of the CROCUS experimental reactor and proposes solutions to the challenges one may face while modeling a research reactor with a complex geometry. Indeed, the double-lattice configuration of CROCUS makes it difficult to use codes for neutron diffusion and transport relying on a structured mesh description. For this reason, and based on the available in-house competences, we decided to make use of the neutronic capabilities of the GeN-Foam multiphysics solver, which takes advantage of general finite volume methodologies on unstructured meshes to provide sufficient flexibility for the study of unconventional reactor designs. In this work, GeN-Foam is used to build a first SP3 model of CROCUS based on an unstructured mesh to have an explicit modeling of the double lattice and the water gap between the two lattices. Form functions are then used to reconstruct the intra-pin fission rates for validation against measured distributions. We also discuss the limitations of the SP3 approximation of neutron transport in regions with steep neutron flux gradients and the planned future developments. Full article
(This article belongs to the Special Issue State-of-Art in Nuclear Reactor Physics)
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12 pages, 682 KiB  
Article
Current Status and On-Going Development of VTT’s Kraken Core Physics Computational Framework
by Jaakko Leppänen, Ville Valtavirta, Antti Rintala, Ville Hovi, Riku Tuominen, Jussi Peltonen, Markus Hirvensalo, Eric Dorval, Unna Lauranto and Rebekka Komu
Energies 2022, 15(3), 876; https://doi.org/10.3390/en15030876 - 25 Jan 2022
Cited by 13 | Viewed by 4401
Abstract
The Kraken computational framework is a new modular calculation system designed for coupled core physics calculations. The development started at VTT Technical Research Centre of Finland in 2017, with the aim to replace VTT’s outdated legacy codes used for the deterministic safety analyses [...] Read more.
The Kraken computational framework is a new modular calculation system designed for coupled core physics calculations. The development started at VTT Technical Research Centre of Finland in 2017, with the aim to replace VTT’s outdated legacy codes used for the deterministic safety analyses of Finnish power reactors. In addition to conventional large PWRs and BWRs, Kraken is intended to be used for the modeling of SMRs and emerging non-LWR technologies. The main computational modules include the Serpent Monte Carlo neutron and photon transport code, the Ants nodal neutronics solver, the FINIX fuel behavior module and the Kharon thermal hydraulics code, all developed at VTT. The core physics solution can be further coupled to system-scale simulations. In addition to development, significant effort has been devoted to verification and validation of the implemented methodologies. The reduced-order Ants code has been successfully used for steady-state, transient and burnup simulations of PWRs with rectangular and hexagonal core geometry. The Ants–Kharon–FINIX code sequence is actively used for the core design tasks in VTT’s district heating reactor project. This paper is a general overview on the background, functional description, current status and future plans for the Kraken framework. Due to the short history of development, Kraken has not yet been comprehensively validated or applied to full-scale core physics calculations. A review of previous studies is instead provided to exemplify the practical use. Full article
(This article belongs to the Special Issue State-of-Art in Nuclear Reactor Physics)
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37 pages, 9423 KiB  
Article
Overview of the Tolerance Limit Calculations with Application to TSURFER
by Hany Abdel-Khalik, Dongli Huang, Ugur Mertyurek, William Marshall and William Wieselquist
Energies 2021, 14(21), 7092; https://doi.org/10.3390/en14217092 - 29 Oct 2021
Cited by 3 | Viewed by 3255
Abstract
To establish confidence in the results of computerized physics models, a key regulatory requirement is to develop a scientifically defendable process. The methods employed for confidence, characterization, and consolidation, or C3, are statistically involved and are often accessible only to [...] Read more.
To establish confidence in the results of computerized physics models, a key regulatory requirement is to develop a scientifically defendable process. The methods employed for confidence, characterization, and consolidation, or C3, are statistically involved and are often accessible only to avid statisticians. This manuscript serves as a pedagogical presentation of the C3 process to all stakeholders—including researchers, industrial practitioners, and regulators—to impart an intuitive understanding of the key concepts and mathematical methods entailed by C3. The primary focus is on calculation of tolerance limits, which is the overall goal of the C3 process. Tolerance limits encode the confidence in the calculation results as communicated to the regulator. Understanding the C3 process is especially critical today, as the nuclear industry is considering more innovative ways to assess new technologies, including new reactor and fuel concepts, via an integrated approach that optimally combines modeling and simulation and minimal targeted validation experiments. This manuscript employs intuitive, analytical, numerical, and visual representations to explain how tolerance limits may be calculated for a wide range of configurations, and it also describes how their values may be interpreted. Various verification tests have been developed to test the calculated tolerance limits and to help delineate their values. The manuscript demonstrates the calculation of tolerance limits for TSURFER, a computer code developed by the Oak Ridge National Laboratory for criticality safety applications. The goal is to evaluate the tolerance limit for TSURFER-determined criticality biases to support the determination of upper, subcritical limits for regulatory purposes. Full article
(This article belongs to the Special Issue State-of-Art in Nuclear Reactor Physics)
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