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 23.5 days after submission; acceptance to publication is undertaken in 15.5 days (median values for papers published in this journal in the second half of 2023).
- Recognition of Reviewers: APC discount vouchers, optional signed peer review, and reviewer names published annually in the journal.
Latest Articles
Neutron Yield Predictions with Artificial Neural Networks: A Predictive Modeling Approach
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.
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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
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Vladimir Chakin, Rolf Rolli, Ramil Gaisin and Wouter van Renterghem
J. Nucl. Eng. 2024, 5(1), 111-113; https://doi.org/10.3390/jne5010008 - 08 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
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Jean-Luc Autran and Daniela Munteanu
J. Nucl. Eng. 2024, 5(1), 91-110; https://doi.org/10.3390/jne5010007 - 05 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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Open AccessArticle
Design and Application of DG-FEM Basis Functions for Neutron Transport on Two-Dimensional and Three-Dimensional Hexagonal Meshes
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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
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Doğaç Tarı, Teodora Retegan Vollmer and Christine Geers
J. Nucl. Eng. 2024, 5(1), 57-73; https://doi.org/10.3390/jne5010005 - 06 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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Open AccessTechnical Note
Burnup-Dependent Neutron Spectrum Behaviour of a Pressurised Water Reactor Fuel Assembly
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Bright Madinka Mweetwa and Marat Margulis
J. Nucl. Eng. 2024, 5(1), 44-56; https://doi.org/10.3390/jne5010004 - 29 Jan 2024
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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
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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
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Tracey Spoerer, Robert Schickler and Steven Reese
J. Nucl. Eng. 2024, 5(1), 13-25; https://doi.org/10.3390/jne5010002 - 03 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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Open AccessArticle
Research on the Influence of Negative KERMA Factors on the Power Distribution of a Lead-Cooled Fast Reactor
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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
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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
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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.
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(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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Open AccessArticle
Application of Machine Learning for Classification of Nuclear Reactor Operational Status Using Magnetic Field Sensors
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Braden Burt, Brett J. Borghetti, Anthony Franz, Darren Holland and Abigail Bickley
J. Nucl. Eng. 2023, 4(4), 723-731; https://doi.org/10.3390/jne4040045 - 06 Dec 2023
Abstract
The nuclear fuel cycle forms the basis for producing special nuclear materials used in nuclear weapons via a series of interdependent industrial operations. These industrial operations each produce characteristic emanations that can be gathered to ascertain signatures of facility operations. Machine learning and
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The nuclear fuel cycle forms the basis for producing special nuclear materials used in nuclear weapons via a series of interdependent industrial operations. These industrial operations each produce characteristic emanations that can be gathered to ascertain signatures of facility operations. Machine learning and deep learning techniques were applied to time series magnetic field sensor data collected at the High Flux Isotope Reactor (HFIR) to assess the feasibility of determining the ON/OFF operational state of the reactor. When data collected by the sensor near the cooling fans, position 9, are transformed to the frequency domain, it was found that both machine and deep learning methods were able to classify the operational state of the reactor with a balanced accuracy of over 90%. This result suggests that the utilized methods show promise for application as techniques to verify declared activities involving nuclear reactors. Additional effort is recommended to develop models and architectures that will more fully capitalize on the data’s temporal nature by incorporating the magnetic field’s time dependence to improve the model’s robustness and classification performance.
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(This article belongs to the Special Issue Nuclear Security and Nonproliferation Research and Development)
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Open AccessCommunication
Oxidation of Alloy X-750 with Low Iron Content in Simulated BWR Environment
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Silvia Tuzi, Krystyna Stiller and Mattias Thuvander
J. Nucl. Eng. 2023, 4(4), 711-722; https://doi.org/10.3390/jne4040044 - 29 Nov 2023
Cited by 1
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This paper presents an investigation of the oxidation of Alloy X-750 containing 5 wt% iron in a simulated boiling water reactor (BWR) environment. The specimens were exposed by a water jet (10 m/s) at 286 °C for durations ranging from 2 to 840
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This paper presents an investigation of the oxidation of Alloy X-750 containing 5 wt% iron in a simulated boiling water reactor (BWR) environment. The specimens were exposed by a water jet (10 m/s) at 286 °C for durations ranging from 2 to 840 h, and the development of the oxide microstructure was mainly studied using electron microscopy. The results showed that the oxide scale consists of blocky crystals of trevorite on top of a porous inner layer rich in Ni and Cr. After the longest exposure time, the trevorite crystals completely covered the specimen surface. The study further revealed that the rate at which the oxide grew and the metal dissolved both decreased with time, and the metal thinning process appeared to be sub-parabolic. Given the significant variation in iron content in the X-750 specification, the influence of this element on the material’s corrosion performance in BWR was examined by comparing the results from this investigation with those from previous work on material containing 8 wt% Fe. The study demonstrates that the oxide growth, metal dissolution and metal thinning were slower in the material with a higher iron content, indicating the importance of this element in limiting the degradation of Alloy X-750 in BWR environments.
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Open AccessArticle
Estimation of Continuous Distribution of Iterated Fission Probability Using an Artificial Neural Network with Monte Carlo-Based Training Data
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Delgersaikhan Tuya and Yasunobu Nagaya
J. Nucl. Eng. 2023, 4(4), 691-710; https://doi.org/10.3390/jne4040043 - 06 Nov 2023
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,
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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.
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(This article belongs to the Special Issue Monte Carlo Simulation in Reactor Physics)
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Open AccessReview
Flow Characterisation Using Fibre Bragg Gratings and Their Potential Use in Nuclear Thermal Hydraulics Experiments
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Harvey Oliver Plows, Jinfeng Li, Marcus Dahlfors and Marat Margulis
J. Nucl. Eng. 2023, 4(4), 668-690; https://doi.org/10.3390/jne4040042 - 25 Oct 2023
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With the ever-increasing role that nuclear power is playing to meet the aim of net zero carbon emissions, there is an intensified demand for understanding the thermal hydraulic phenomena at the heart of current and future reactor concepts. In response to this demand,
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With the ever-increasing role that nuclear power is playing to meet the aim of net zero carbon emissions, there is an intensified demand for understanding the thermal hydraulic phenomena at the heart of current and future reactor concepts. In response to this demand, the development of high-resolution flow analysis instrumentation is of increased importance. One such under-utilised and under-researched instrumentation technology, in the context of fluid flow analysis, is fibre Bragg grating (FBG)-based sensors. This technology allows for the construction of simple, minimally invasive instruments that are resistant to high temperatures, high pressures and corrosion, while being adaptable to measure a wide range of fluid properties, including temperature, pressure, refractive index, chemical concentration, flow rate and void fraction—even in opaque media. Furthermore, concertinaing FBG arrays have been developed capable of reconstructing 3D images of large phase structures, such as bubbles in slug flow, that interact with the array. Currently a significantly under-explored application, FBG-based instrumentation thus shows great potential for utilisation in experimental thermal hydraulics; expanding the available flow characterisation and imaging technologies. Therefore, this paper will present an overview of current FBG-based flow characterisation technologies, alongside a systematic review of how these techniques have been utilised in nuclear thermal hydraulics experiments. Finally, a discussion will be presented regarding how these techniques can be further developed and used in nuclear research.
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Open AccessArticle
A Consistent One-Dimensional Multigroup Diffusion Model for Molten Salt Reactor Neutronics Calculations
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Mohamed Elhareef, Zeyun Wu and Massimiliano Fratoni
J. Nucl. Eng. 2023, 4(4), 654-667; https://doi.org/10.3390/jne4040041 - 06 Oct 2023
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Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup
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Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis.
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Open AccessArticle
The Peculiarities of the German Uranium Project (1939–1945)
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Manfred Popp and Piet de Klerk
J. Nucl. Eng. 2023, 4(3), 634-653; https://doi.org/10.3390/jne4030040 - 13 Sep 2023
Abstract
An analysis of the peculiarities of the German Uranium Project (1939–1945) reveals that it was, in many ways, different from what one would expect. There was no work at all on a possible bomb, nor on plutonium. The reactor experiments were limited to
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An analysis of the peculiarities of the German Uranium Project (1939–1945) reveals that it was, in many ways, different from what one would expect. There was no work at all on a possible bomb, nor on plutonium. The reactor experiments were limited to subcritical systems and did not attempt to achieve the proclaimed goal of a self-sustaining chain reaction. The so-far identified deficits (lack of interest in Nazi circles, mismanagement, scientific mistakes, and deteriorating work conditions during the war) are relevant but not sufficient for explaining the peculiarities. We deduce that the scientists involved, and even the Heereswaffenamt (army ordnance), shied away from making progress, not only towards a bomb but even towards a reactor. They did not fail; they rather renounced a possible success in order not to provoke political interest in the development of a bomb.
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Open AccessArticle
Feasibility Study on Production of High-Purity Rhenium-185 by Nuclear Transmutation of Natural Tantalum
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Yuki Tanoue, Tsugio Yokoyama and Masaki Ozawa
J. Nucl. Eng. 2023, 4(3), 625-633; https://doi.org/10.3390/jne4030039 - 01 Sep 2023
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Rhenium-186 (Re-186) has attracted attention as a medical isotope. The feasibility of producing Re-185, the raw material for Re-186, using a fast reactor was evaluated using a continuous energy Monte Carlo code. The irradiation of natural tantalum (Ta) in the fast reactor can
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Rhenium-186 (Re-186) has attracted attention as a medical isotope. The feasibility of producing Re-185, the raw material for Re-186, using a fast reactor was evaluated using a continuous energy Monte Carlo code. The irradiation of natural tantalum (Ta) in the fast reactor can produce Re-185 with an isotopic purity of 99%. A two-step irradiation process with different moderators was found to improve the production rate of Re-185. Specifically, this can be achieved by using zirconium hydride (ZrH1.7) as a moderator in the first transmutation process from natural Ta to tungsten (W), and then zirconium deuteride (ZrD1.7) as a moderator in the second transmutation process from W to Re-185. Due to the two-step irradiation, the production rate of Re-185 from Ta can be increased up to a maximum of 470 times compared with irradiation without a moderator, and 2.3 g of Re-185 can be obtained from 1571 g of Ta in 1 year of irradiation. The proposed isotope production method is a new method that is different from the conventional electromagnetic enrichment process.
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Open AccessReview
A Review of Candidates for a Validation Data Set for High-Assay Low-Enrichment Uranium Fuels
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Mark D. DeHart, John Darrell Bess and Germina Ilas
J. Nucl. Eng. 2023, 4(3), 602-624; https://doi.org/10.3390/jne4030038 - 16 Aug 2023
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Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few
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Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and software-validation efforts. This paper reviews the available evaluated HALEU fuel benchmarks and some of the potentially relevant benchmarks for fresh highly enriched uranium. It then introduces experimental data for HALEU fuel irradiated at Idaho National Laboratory, from relatively recent irradiation programs at the Advanced Test Reactor. Such data should be evaluated and, if valuable, collected into detailed benchmark specifications to meet the needs of HALEU-based reactor designers.
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Open AccessArticle
Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities
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Francisco A. Hernández, Pietro Arena, Lorenzo V. Boccaccini, Ion Cristescu, Alessandro Del Nevo, Pierre Sardain, Gandolfo A. Spagnuolo, Marco Utili, Alessandro Venturini and Guangming Zhou
J. Nucl. Eng. 2023, 4(3), 565-601; https://doi.org/10.3390/jne4030037 - 03 Aug 2023
Cited by 5
Abstract
The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion
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The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.
Full article
(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)
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Open AccessArticle
Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor
by
Vladimir Chakin, Rolf Rolli, Ramil Gaisin and Wouter van Renterghem
J. Nucl. Eng. 2023, 4(3), 552-564; https://doi.org/10.3390/jne4030036 - 02 Aug 2023
Cited by 1
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The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests
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The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.
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Open AccessArticle
The Plutonium Temperature Effect Program
by
Nicolas Leclaire and Vaibhav Jaiswal
J. Nucl. Eng. 2023, 4(3), 535-551; https://doi.org/10.3390/jne4030035 - 02 Aug 2023
Cited by 1
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Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that
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Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.
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