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Design and Application of DG-FEM Basis Functions for Neutron Transport on Two-Dimensional and Three-Dimensional Hexagonal Meshes
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Interactions of Low-Energy Muons with Silicon: Numerical Simulation of Negative Muon Capture and Prospects for Soft Errors
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Burnup-Dependent Neutron Spectrum Behaviour of a Pressurised Water Reactor Fuel Assembly
Journal Description
Journal of Nuclear Engineering
Journal of Nuclear Engineering
is an international, peer-reviewed, open access journal on nuclear and radiation sciences and applications, published quarterly online by MDPI.
- Open Access— free for readers, with article processing charges (APC) paid by authors or their institutions.
- High Visibility: indexed within ESCI (Web of Science), EBSCO and other databases.
- Rapid Publication: manuscripts are peer-reviewed and a first decision is provided to authors approximately 34.3 days after submission; acceptance to publication is undertaken in 9.5 days (median values for papers published in this journal in the first half of 2024).
- Recognition of Reviewers: APC discount vouchers, optional signed peer review, and reviewer names published annually in the journal.
Latest Articles
Mini-Reactor Proliferation-Resistant Fuel with Burnable Gadolinia in Once-Through Operation Cycle Performance Verification
J. Nucl. Eng. 2024, 5(3), 318-329; https://doi.org/10.3390/jne5030021 - 28 Aug 2024
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A miniature nuclear reactor is desirable for deployment as a localized nuclear power station in support of a carbon-free power supply. Coupling aspects of proliferation-resistant fuel with natural burnable absorber loading are evaluated for once-through operation cycle performance to minimize the need for
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A miniature nuclear reactor is desirable for deployment as a localized nuclear power station in support of a carbon-free power supply. Coupling aspects of proliferation-resistant fuel with natural burnable absorber loading are evaluated for once-through operation cycle performance to minimize the need for refueling and fuel shuffling operations. The incorporation of 0.075 wt.% 237Np provides favorable plutonium isotopic vectors throughout an operational lifetime of 5.5 years. providing 35 MWe. Core performance was assessed using a verification-by-comparison approach for core designs with or without 237Np and/or gadolinia burnable absorber. Burnup Monte Carlo calculations were performed via MCOS coupling of MCNP and ORIGEN to an achievable burnup of ~62.5 GWd/t. The results demonstrate a minimal penalty to reactor performance due to the addition of these materials as compared against the reference design. Coupling of a proliferation-resistant fuel concept with a uniform loading of natural gadolinia burnable absorber for LEU+ fuel (7.5 wt.% 235U/U UO2) provides favorable excess reactivity considerations with minimized concerns for additional residual waste and more uniform distribution of un-depleted 235U in discharged fuel assemblies.
Full article
Open AccessReview
Trends and Perspectives on Nuclear Waste Management: Recovering, Recycling, and Reusing
by
Maria Letizia Terranova and Odilon A. P. Tavares
J. Nucl. Eng. 2024, 5(3), 299-317; https://doi.org/10.3390/jne5030020 - 13 Aug 2024
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This paper focuses on the highly radioactive, long-lasting nuclear waste produced by the currently operating fission reactors and on the sensitive issue of spent fuel reprocessing. Also included is a short description of the fission process and a detailed analysis of the more
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This paper focuses on the highly radioactive, long-lasting nuclear waste produced by the currently operating fission reactors and on the sensitive issue of spent fuel reprocessing. Also included is a short description of the fission process and a detailed analysis of the more hazardous radioisotopes produced either by secondary reactions occurring in the nuclear installations or by decay of the fission fragments. The review provides an overview of the strategies presently adopted to minimize the harmfulness of the nuclear waste to be disposed, with a focus on the development and implementation of methodologies for the spent fuel treatments. The partitioning-conditioning and partitioning-transmutation options are analyzed as possible solutions to decrease the presence of long-lived highly radioactive isotopes. Also discussed are the chemical/physical approaches proposed for the recycling of the spent fuel and for the reusing of some technologically relevant isotopes in industrial and pharmaceutical areas. A brief indication is given of the opportunities offered by innovative types of reactors and/or of new fuel cycles to solve the issues presently associated with radioactive waste.
Full article
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Open AccessArticle
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
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
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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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Open AccessArticle
Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor
by
Kang Seog Kim, Andrew Ward, Ugur Mertyurek, Mehdi Asgari and William Wieselquist
J. Nucl. Eng. 2024, 5(3), 260-273; https://doi.org/10.3390/jne5030018 - 1 Aug 2024
Abstract
The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by
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The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.
Full article
(This article belongs to the Special Issue Validation of Code Packages for Light Water Reactor Physics Analysis)
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Open AccessArticle
Validation of the SCALE/Polaris−PARCS Code Procedure with the ENDF/B-VII.1 AMPX 56-Group Library: Pressurized Water Reactor
by
Kang Seog Kim, Byoung-Kyu Jeon, Andrew Ward, Ugur Mertyurek, Matthew Jessee and William Wieselquist
J. Nucl. Eng. 2024, 5(3), 246-259; https://doi.org/10.3390/jne5030017 - 23 Jul 2024
Abstract
This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of
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This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors.
Full article
(This article belongs to the Special Issue Validation of Code Packages for Light Water Reactor Physics Analysis)
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Open AccessArticle
Phenomenological Nondimensional Parameter Decomposition to Enhance the Use of Simulation Modeling in Fire Probabilistic Risk Assessment of Nuclear Power Plants
by
Sari Alkhatib, Tatsuya Sakurahara, Seyed Reihani, Ernest Kee, Brian Ratte, Kristin Kaspar, Sean Hunt and Zahra Mohaghegh
J. Nucl. Eng. 2024, 5(3), 226-245; https://doi.org/10.3390/jne5030016 - 2 Jul 2024
Abstract
Simulation modeling is crucial in support of probabilistic risk assessment (PRA) for nuclear power plants (NPPs). There is a challenge, however, associated with simulation modeling that relates to the time and resources required for collecting data to determine the values of the input
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Simulation modeling is crucial in support of probabilistic risk assessment (PRA) for nuclear power plants (NPPs). There is a challenge, however, associated with simulation modeling that relates to the time and resources required for collecting data to determine the values of the input parameters. To alleviate this challenge, this article develops a formalized methodology to generate surrogate values of input parameters grounded on the decomposition of phenomenological nondimensional parameters (PNPs) while avoiding detailed data collection. While the fundamental principles of the proposed methodology can be applicable to various hazards, the developments in this article focus on fire PRA as an example application area for which resource intensiveness is recognized as a practical challenge. This article also develops a computational platform to automate the PNP decomposition and seamlessly integrates it with state-of-practice fire scenario analysis. The applicability of the computational platform is demonstrated through a multi-compartment fire case study at an NPP. The computational platform, with its embedded PNP decomposition methodology, can substantially reduce the effort required for input data collection and extraction, thereby facilitating the efficient use of simulation modeling in PRA and enhancing the fire scenario screening analysis.
Full article
(This article belongs to the Special Issue Reliability Analysis and Risk Assessment of Nuclear Systems)
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Open AccessArticle
Reinforcement Learning-Based Control Sequence Optimization for Advanced Reactors
by
Khang H. N. Nguyen, Andy Rivas, Gregory Kyriakos Delipei and Jason Hou
J. Nucl. Eng. 2024, 5(3), 209-225; https://doi.org/10.3390/jne5030015 - 1 Jul 2024
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The last decade has seen the development and application of data-driven methods taking off in nuclear engineering research, aiming to improve the safety and reliability of nuclear power. This work focuses on developing a reinforcement learning-based control sequence optimization framework for advanced nuclear
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The last decade has seen the development and application of data-driven methods taking off in nuclear engineering research, aiming to improve the safety and reliability of nuclear power. This work focuses on developing a reinforcement learning-based control sequence optimization framework for advanced nuclear systems, which not only aims to enhance flexible operations, promoting the economics of advanced nuclear technology, but also prioritizing safety during normal operation. At its core, the framework allows the sequence of operational actions to be learned and optimized by an agent to facilitate smooth transitions between the modes of operations (i.e., load-following), while ensuring that all safety significant system parameters remain within their respective limits. To generate dynamic system responses, facilitate control strategy development, and demonstrate the effectiveness of the framework, a simulation environment of a pebble-bed high-temperature gas-cooled reactor was utilized. The soft actor-critic algorithm was adopted to train a reinforcement learning agent, which can generate control sequences to maneuver plant power output in the range between 100% and 50% of the nameplate power through sufficient training. It was shown in the performance validation that the agent successfully generated control actions that maintained electrical output within a tight tolerance of 0.5% from the demand while satisfying all safety constraints. During the mode transition, the agent can maintain the reactor outlet temperature within °C and steam pressure within 0.1 MPa of their setpoints, respectively, by dynamically adjusting control rod positions, control valve openings, and pump speeds. The results demonstrate the effectiveness of the optimization framework and the feasibility of reinforcement learning in designing control strategies for advanced reactor systems.
Full article
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Open AccessCommunication
New Mini Neutron Tubes with Multiple Applications
by
Ka-Ngo Leung
J. Nucl. Eng. 2024, 5(3), 197-208; https://doi.org/10.3390/jne5030014 - 26 Jun 2024
Cited by 1
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Recent experimental investigations have demonstrated that a substantial amount of H−/D− ions can be formed by thermal desorption processes. Based on these new findings, new mini axial and coaxial-type neutron tubes have been developed for the production of high or
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Recent experimental investigations have demonstrated that a substantial amount of H−/D− ions can be formed by thermal desorption processes. Based on these new findings, new mini axial and coaxial-type neutron tubes have been developed for the production of high or low-energy neutrons via the d-d, d-10B, d-7Li or p-7Li nuclear reactions. By operating these mini neutron tubes with a high frequency AC high-voltage supply, short pulses of high intensity neutron beams can be generated. Multiple applications, such as carbon and well logging, neutron imaging, cancer therapy, medical isotope production, fission reactor start-up, fusion reactor material evaluation, homeland security and space exploration can be performed with the subcompact neutron generator system. It is shown that the performance of these new mini neutron tubes can exceed those of the conventional plasma-based neutron sources.
Full article
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Open AccessTechnical Note
Immediate Radiological Risk Evaluation after a Hypothetical Radioactive Off-Site Release Event
by
Ana Carolina Lodi Lobato, Sérgio Gavazza, Avelino Santos, Rodrigo Carneiro Curzio and Edson R. Andrade
J. Nucl. Eng. 2024, 5(2), 186-196; https://doi.org/10.3390/jne5020013 - 19 Jun 2024
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This study used an analytical computational model to evaluate safe zones in contaminated areas that may result from a hypothetical significant off-site release from a nuclear power plant. The model, considering local atmospheric stability, wind direction, and location, calculates the expected total effective
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This study used an analytical computational model to evaluate safe zones in contaminated areas that may result from a hypothetical significant off-site release from a nuclear power plant. The model, considering local atmospheric stability, wind direction, and location, calculates the expected total effective dose equivalent (TEDE) and potential safety zones. This research, focused on an area near a nuclear facility affected by an accidental release, used SCALE and HotSpot Health Physics codes to simulate the reactor’s core inventory content and off-site release. This study’s findings underscore that the risk of developing solid cancer (testing morbidity) is influenced by both local atmospheric stability and the composition of the potentially affected population (primarily age and sex). These findings, backed by an analytical approach, can significantly influence logistical and operational planning. The utilization of computer simulations can also aid in creating flexible response scenarios to real events.
Full article
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Open AccessArticle
Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor
by
Xuzhong Kang, Guifeng Zhu, Jianhui Wu, Rui Yan, Yang Zou and Yafen Liu
J. Nucl. Eng. 2024, 5(2), 168-185; https://doi.org/10.3390/jne5020012 - 10 May 2024
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The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan.
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The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.
Full article
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Open AccessArticle
The Effects of Irradiation on Structure and Leaching of Pure and Doped Thin-Film Ceria SIMFUEL Models Prepared via Polymer-Templated Deposition
by
Alistair F. Holdsworth, Zizhen Feng, Ruth Edge, John P. Waters, Alice M. Halman, David Collison, Kathryn George, Louise S. Natrajan and Melissa A. Denecke
J. Nucl. Eng. 2024, 5(2), 150-167; https://doi.org/10.3390/jne5020011 - 8 May 2024
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When studying hazardous materials such as spent nuclear fuel (SNF), the minimisation of sample volumes is essential, together with the use of chemically-similar surrogates where possible. For example, the bulk behaviour of urania (UO2) can be mimicked by appropriately-engineered thin films
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When studying hazardous materials such as spent nuclear fuel (SNF), the minimisation of sample volumes is essential, together with the use of chemically-similar surrogates where possible. For example, the bulk behaviour of urania (UO2) can be mimicked by appropriately-engineered thin films of sufficient thickness, and inactive materials such as ceria (CeO2) can be used to study the effects within radioactive systems used to fuel nuclear fission. However, thin film properties are sensitive to the preparative method, many of which require the use of highly toxic precursors and specialised apparatus (e.g., chemical vapour deposition). To address this, we present the development of a flexible, tuneable, scalable method for the preparation of thin-film CeO2 SIMFUEL models with a thickness of ≈5 μm. The effects of γ irradiation (up to 100 kGy) and dopants including trivalent lanthanides (Ln3+) and simulant ε-particles on the structure and long-term leaching of these systems under SNF storage conditions were explored, alongside the context of this within further work. It was found that the sensitivity of CeO2 films to reduction upon irradiation, particularly in the presence of simulant ε-particles, resulted in increased leaching of Ce (as CeIII), while trivalent lanthanides (Nd3+ and Eu3+) had a minimal effect on Ce leaching.
Full article
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Open AccessArticle
Numerical Investigation of Butterfly Valve Performance in Variable Valve Sizes, Positions and Flow Regimes
by
Anutam Bairagi, Mingfu He and Minghui Chen
J. Nucl. Eng. 2024, 5(2), 128-149; https://doi.org/10.3390/jne5020010 - 24 Apr 2024
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Reliability and efficiency of valves are necessary for precise control and sufficient heat-flow to heat application plants for the integrated energy systems of nuclear power plants (NPPs). Strategic Management Analysis Requirement and Technology (SMART) valves’ ability to control flow and assess environmental parameters
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Reliability and efficiency of valves are necessary for precise control and sufficient heat-flow to heat application plants for the integrated energy systems of nuclear power plants (NPPs). Strategic Management Analysis Requirement and Technology (SMART) valves’ ability to control flow and assess environmental parameters stands out for these requirements. Their ability to sustain the downstream flow rate, prevent reverse flow, and maintain pressure in the heat transport loop is much more efficient with the integration of sensors and intelligent algorithms. For assessing valve performance and monitoring, mechanical design and operating conditions are two important parameters. In this study, the butterfly valves of three different sizes are simulated with water and steam using STAR-CCM+ in various flow regimes and positions to analyze performance parameters to strategize an automated control system for efficiently balancing the heat–transport network. Also, flow behavior is studied using velocity and pressure fields for valve–body geometry optimization. It can be observed, through performance parameters, that the valves are suitable for operation between 30° and 90° positions with significantly low loss coefficients and high flow coefficients, and the performance parameters follow a certain pattern in both water and steam flow in each scenario.
Full article
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Open AccessArticle
Neutron Yield Predictions with Artificial Neural Networks: A Predictive Modeling Approach
by
Benedikt Schmitz and Stefan Scheuren
J. Nucl. Eng. 2024, 5(2), 114-127; https://doi.org/10.3390/jne5020009 - 31 Mar 2024
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The development of compact neutron sources for applications is extensive and features many approaches. For ion-based approaches, several projects with different parameters exist. This article focuses on ion-based neutron production below the spallation barrier for proton and deuteron beams with arbitrary energy distributions
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The development of compact neutron sources for applications is extensive and features many approaches. For ion-based approaches, several projects with different parameters exist. This article focuses on ion-based neutron production below the spallation barrier for proton and deuteron beams with arbitrary energy distributions with kinetic energies from 3 to 97 . This model makes it possible to compare different ion-based neutron source concepts against each other quickly. This contribution derives a predictive model using Monte Carlo simulations (an order of 50,000 simulations) and deep neural networks. It is the first time a model of this kind has been developed. With this model, lengthy Monte Carlo simulations, which individually take a long time to complete, can be circumvented. A prediction of neutron spectra then takes some milliseconds, which enables fast optimization and comparison. The models’ shortcomings for low-energy neutrons (< ) and the cut-off prediction uncertainty ( ) are addressed, and mitigation strategies are proposed.
Full article
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Open AccessCorrection
Correction: Chakin et al. Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor. J. Nucl. Eng. 2023, 4, 552–564
by
Vladimir Chakin, Rolf Rolli, Ramil Gaisin and Wouter van Renterghem
J. Nucl. Eng. 2024, 5(1), 111-113; https://doi.org/10.3390/jne5010008 - 8 Mar 2024
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The authors would like to make the following corrections to the published paper [...]
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Open AccessArticle
Interactions of Low-Energy Muons with Silicon: Numerical Simulation of Negative Muon Capture and Prospects for Soft Errors
by
Jean-Luc Autran and Daniela Munteanu
J. Nucl. Eng. 2024, 5(1), 91-110; https://doi.org/10.3390/jne5010007 - 5 Mar 2024
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In this paper, the interactions of low-energy muons (E < 10 MeV) with natural silicon, the basic material of microelectronics, are studied by Geant4 and SRIM simulation. The study is circumscribed to muons susceptible to slowdown/stop in the target and able to transfer
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In this paper, the interactions of low-energy muons (E < 10 MeV) with natural silicon, the basic material of microelectronics, are studied by Geant4 and SRIM simulation. The study is circumscribed to muons susceptible to slowdown/stop in the target and able to transfer sufficient energy to the semiconductor to create single events in silicon devices or related circuits. The capture of negative muons by silicon atoms is of particular interest, as the resulting nucleus evaporation and its effects can be catastrophic in terms of the emission of secondary ionizing particles ranging from protons to aluminum ions. We investigate in detail these different nuclear capture reactions in silicon and quantitatively evaluate their relative importance in terms of number of products, energy, linear energy transfer, and range distributions, as well as in terms of charge creation in silicon. Finally, consequences in the domain of soft errors in microelectronics are discussed.
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Graphical abstract
Open AccessArticle
Design and Application of DG-FEM Basis Functions for Neutron Transport on Two-Dimensional and Three-Dimensional Hexagonal Meshes
by
Ansar Calloo, David Labeurthre and Romain Le Tellier
J. Nucl. Eng. 2024, 5(1), 74-90; https://doi.org/10.3390/jne5010006 - 26 Feb 2024
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Reactor design requires safety studies to ensure that the reactors will behave appropriately under incidental or accidental situations. Safety studies often involve multiphysics simulations where several branches of reactor physics are necessary to model a given phenomenon. In those situations, it has been
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Reactor design requires safety studies to ensure that the reactors will behave appropriately under incidental or accidental situations. Safety studies often involve multiphysics simulations where several branches of reactor physics are necessary to model a given phenomenon. In those situations, it has been observed that the neutron transport part is still a bottleneck in terms of computational times, with more than 80% of the total time. In the case of hexagonal lattice reactors, transport solvers usually invert the discretised Boltzmann equation by discretising the regular hexagon into lozenges or triangles. In this work, we seek to reduce the computational burden of the neutron transport solver by designing a numerical spatial discretisation scheme that would be more appropriate for honeycomb meshes. In our past research efforts, we have set up interesting discretisation schemes in the finite element setting in 2D, and we wish to extend them to 3D geometries that are prisms with a hexagonal base. In 3D, a rigorous method was derived to shrink the tensor product between 2D and 1D bases to minimum terms. We have applied these functions successfully on a reactor benchmark—Takeda Model 4—to compare and contrast the numerical results in a physical setting.
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Open AccessArticle
Reaction Capsule Design for Interaction of Heavy Liquid Metal Coolant, Fuel Cladding, and Simulated JOG Phase at Accident Conditions
by
Doğaç Tarı, Teodora Retegan Vollmer and Christine Geers
J. Nucl. Eng. 2024, 5(1), 57-73; https://doi.org/10.3390/jne5010005 - 6 Feb 2024
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High temperature corrosion of fuel cladding material (15-15Ti) in high burn-up situations has been an important topic for molten metal-cooled Gen-IV reactors. The present study aims to investigate the simultaneous impact of liquid lead (coolant side) and cesium molybdate (fuel side) on the
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High temperature corrosion of fuel cladding material (15-15Ti) in high burn-up situations has been an important topic for molten metal-cooled Gen-IV reactors. The present study aims to investigate the simultaneous impact of liquid lead (coolant side) and cesium molybdate (fuel side) on the cladding tube material. A capsule was designed and built for experiments between 600 °C and 1000 °C. In order to simulate a cladding breach scenario, a notch design on the cladding tube was investigated pre- and postexposure. Material thinning by corrosion and leaching at temperatures ≥ 900 °C caused breaches at the notches after 168 h exposure. The temperature dependent cladding thinning phenomenon was used for kinetic interpretation. As the first of a two-part study, this paper will focus on the exposure capsule performance, including metallographic cross-section preparation and preliminary results on the interface chemistry.
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Graphical abstract
Open AccessTechnical Note
Burnup-Dependent Neutron Spectrum Behaviour of a Pressurised Water Reactor Fuel Assembly
by
Bright Madinka Mweetwa and Marat Margulis
J. Nucl. Eng. 2024, 5(1), 44-56; https://doi.org/10.3390/jne5010004 - 29 Jan 2024
Cited by 1
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Understanding the behaviour of a neutron spectrum with burnup is important for describing various phenomena associated with reactor operation. The quest to understand the neutron spectrum comes with a lot of questions. One question that is usually asked by students is: Does the
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Understanding the behaviour of a neutron spectrum with burnup is important for describing various phenomena associated with reactor operation. The quest to understand the neutron spectrum comes with a lot of questions. One question that is usually asked by students is: Does the neutron spectrum harden or soften with burnup? Most textbooks used by students do not provide a definite answer to this question. This paper seeks to answer this question using a 3D model of a standard 17 × 17 pressurised water reactor fuel assembly. Two cases were studied using the Serpent Monte Carlo code: the first considered the fuel assembly with constant boron concentration (traditionally found in many published papers), and the second considered boron iteration (where the boron concentration was reduced with burnup). Neutron spectra for the two cases at beginning of life and end of life were compared for spectral shifts. In addition, thermal spectral indices were used to assess spectrum hardening or softening with burnup. Spectral shifts to lower energies were observed in the thermal region of the neutron spectrum, whereas the fast region experienced no spectral shift. There was an increase in thermal spectral indices indicating that the spectrum became soft with burnup.
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Open AccessArticle
Gamma-ray Spectroscopy in Low-Power Nuclear Research Reactors
by
Oskari V. Pakari, Andrew Lucas, Flynn B. Darby, Vincent P. Lamirand, Tessa Maurer, Matthew G. Bisbee, Lei R. Cao, Andreas Pautz and Sara A. Pozzi
J. Nucl. Eng. 2024, 5(1), 26-43; https://doi.org/10.3390/jne5010003 - 26 Jan 2024
Cited by 1
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Gamma-ray spectroscopy is an effective technique for radioactive material characterization, routine inventory verification, nuclear safeguards, health physics, and source search scenarios. Gamma-ray spectrometers typically cannot be operated in the immediate vicinity of nuclear reactors due to their high flux fields and their resulting
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Gamma-ray spectroscopy is an effective technique for radioactive material characterization, routine inventory verification, nuclear safeguards, health physics, and source search scenarios. Gamma-ray spectrometers typically cannot be operated in the immediate vicinity of nuclear reactors due to their high flux fields and their resulting inability to resolve individual pulses. Low-power reactor facilities offer the possibility to study reactor gamma-ray fields, a domain of experiments hitherto poorly explored. In this work, we present gamma-ray spectroscopy experiments performed with various detectors in two reactors: The EPFL zero-power research reactor CROCUS, and the neutron beam facility at the Ohio State University Research Reactor (OSURR). We employed inorganic scintillators (CeBr3), organic scintillators (trans-stilbene and organic glass), and high-purity germanium semiconductors (HPGe) to cover a range of typical—and new—instruments used in gamma-ray spectroscopy. The aim of this study is to provide a guideline for reactor users regarding detector performance, observed responses, and therefore available information in the reactor photon fields up to 2 MeV. The results indicate several future prospects, such as the online (at criticality) monitoring of fission products (like Xe, I, and La), dual-particle sensitive experiments, and code validation opportunities.
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Open AccessArticle
Effects of Neutron Flux Distribution and Control Rod Shadowing on Control Rod Calibrations in the Oregon State TRIGA® Reactor
by
Tracey Spoerer, Robert Schickler and Steven Reese
J. Nucl. Eng. 2024, 5(1), 13-25; https://doi.org/10.3390/jne5010002 - 3 Jan 2024
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Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four
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Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four OSTR control rods is measured using the rod-pull method. Reactor power and period measurements in this method rely on the fission chamber power detector on the north side of the reflector. It is proposed that the location of the fission chamber and the neutron flux distribution in the core may result in an inaccurate reactor period measurement due to the asymmetry of the neutron flux distribution in the OSTR core. The asymmetry of the flux is believed to be more pronounced during super-criticality, resulting in errors in the time-of-power-rise measurements. As a result, control rod calibration experiments may under-predict or over-predict the reactivity worth of certain control rods. A time-independent Monte–Carlo method for the quantification of these effects is presented. Thermal flux maps at the core axial mid-plane are obtained from the model to inform discrepancies between predicted and observed results.
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