Selected Papers from PHYSOR 2020

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

Deadline for manuscript submissions: closed (1 October 2020) | Viewed by 58095

Special Issue Editors


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Guest Editor
Department of Engineering, University of Cambridge, Cambridge CB2 1PZ, UK
Interests: nuclear reactor engineering and design; reactor physics applied to modelling of advanced nuclear energy systems; deterministic and Monte Carlo neutron transport methods with multi-physics coupling

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Guest Editor
Scientific Division Energies, Energies Directorate, CEA Saclay, 91191 Gif-sur-Yvette, France
Interests: research reactors; core physics; simulation and experimental platforms; instrumentation; project management in various international frameworks

Special Issue Information

Dear Colleagues,

Thirty years after the first PHYSOR was held in Marseille, France, the Universities of Cambridge and Birmingham collaborated to host PHYSOR 2020 in the historic city of Cambridge, to celebrate three decades of outstanding research in the field of nuclear reactor physics. The topic of the conference this year was “Transition to a scalable nuclear future”, with the goal of addressing global climate change and the growth of intermittent renewable energy generation. Nuclear energy has been identified by the IPCC as an important component of the energy sector decarbonisation activity, striving for increased flexibility and scalability, improved safety, reduced waste, and—probably most importantly at the moment—cost reduction.

The PHYSOR conference is dedicated to the presenting the latest developments in core physics and design, applied mathematics, nuclear data, model verification and validation, sensitivity analysis and uncertainty quantification. Over 400 papers have been accepted for presentation at the conference, organized in 22 technical tracks, special, and poster sessions. Unfortunately, the global COVID pandemic and official U.K. government restrictions led, at the last moment, to the cancellation of these sessions. However, in recognition of the important contribution to the field of reactor physics and of the hard work done by all of the participants—authors, track chairs, reviewers, and other members of the organising committee, we have decided—as organisers of this important conference—to publish the proceedings in an open access format. In parallel, we have asked the track chairs to nominate a small number of high-quality papers, with unanimous positive reviews, reflecting the spirit of the conference for publication in a Special Issue of this journal dedicated to the PHYSOR 2020 conference. We wish to thank the JNE Founding Editor, Prof. Dan G. Cacuci, for having proposed the creation of this Special Issue.

The overall quality of the conference contributions was outstanding; the papers included in this Special Issue represent the state of the art in research and development for each of the technical tracks.

We hope you will enjoy reading the products of this truly amazing high-quality conference as much as we enjoyed organising it, with the help of the local organising team, as well as the priceless and dedicated help of all of the track leaders and over 230 reviewers from around the globe. We would like to thank you again for your contribution to the PHYSOR2020 conference.

We wish our thriving reactor physics community many more years of exciting and impactful research.

Prof. Dr. Eugene Shwageraus, Physor 2020 General Chair
Dr. Patrick Blaise, Physor 2020 Technical Program Chair
Guest Editors

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Keywords

  • reactor physics
  • mathematics
  • research reactors
  • core design
  • nuclear data
  • fuel cycle

Published Papers (23 papers)

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Research

9 pages, 334 KiB  
Article
The Stability of Linear Diffusion Acceleration Relative to CMFD
by Zackary Dodson, Brendan Kochunas and Edward Larsen
J. Nucl. Eng. 2021, 2(4), 336-344; https://doi.org/10.3390/jne2040027 - 24 Sep 2021
Cited by 3 | Viewed by 2351
Abstract
Coarse Mesh Finite Difference (CMFD) is a widely-used iterative acceleration method for neutron transport problems in which nonlinear terms are introduced in the derivation of the low-order CMFD diffusion equation. These terms, including the homogenized diffusion coefficient, the current coupling coefficients, and the [...] Read more.
Coarse Mesh Finite Difference (CMFD) is a widely-used iterative acceleration method for neutron transport problems in which nonlinear terms are introduced in the derivation of the low-order CMFD diffusion equation. These terms, including the homogenized diffusion coefficient, the current coupling coefficients, and the multiplicative prolongation constant, are subject to numerical instability when a scalar flux estimate becomes sufficiently small or negative. In this paper, we use a suite of contrived problems to demonstrate the susceptibility of CMFD to failure for each of the vulnerable quantities of interest. Our results show that if a scalar flux estimate becomes negative in any portion of phase space, for any iterate, numerical instability can occur. Specifically, the number of outer iterations required for convergence of the CMFD-accelerated transport problem can increase dramatically, or worse, the iteration scheme can diverge. An alternative Linear Diffusion Acceleration (LDA) scheme addresses these issues by explicitly avoiding local nonlinearities. Our numerical results show that the rapid convergence of LDA is unaffected by the very small or negative scalar flux estimates that can adversely affect the performance of CMFD. Therefore, our results demonstrate that LDA is a robust alternative to CMFD for certain sensitive problems in which CMFD can exhibit reduced effectiveness or failure. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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9 pages, 6638 KiB  
Article
Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) Using CTF
by Samuel A. Walker, Abdalla Abou-Jaoude, Zack Taylor, Robert K. Salko and Wei Ji
J. Nucl. Eng. 2021, 2(3), 309-317; https://doi.org/10.3390/jne2030025 - 24 Aug 2021
Cited by 2 | Viewed by 2446
Abstract
With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho [...] Read more.
With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory to be placed in the Advanced Test Reactor (ATR). One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this has on the fuel-salt properties, source term generation, and corrosion control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system, and noble metals (Mo, Tc, Te, etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in CTF (Coolant-Boiling in Rod Arrays—Two Fluids) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design. A noble metal–helium bubble mass transfer model is coupled with the thermal-hydraulic results from CTF to determine the effectiveness of this insoluble fission product (IFP) extraction method for VESIL. Amounts of IFP species extracted to the off-gas system and species distributions in VESIL after a 60-day ATR cycle are reported. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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7 pages, 354 KiB  
Article
Arbitrary-Order Bernstein–Bézier Functions for DGFEM Transport on 3D Polygonal Grids
by Michael Hackemack
J. Nucl. Eng. 2021, 2(3), 239-245; https://doi.org/10.3390/jne2030022 - 31 Jul 2021
Viewed by 2140
Abstract
In this paper, we present an arbitrary-order discontinuous Galerkin finite element discretization of the SN transport equation on 3D extruded polygonal prisms. Basis functions are formed by the tensor product of 2D polygonal Bernstein–Bézier functions and 1D Lagrange polynomials. For a polynomial [...] Read more.
In this paper, we present an arbitrary-order discontinuous Galerkin finite element discretization of the SN transport equation on 3D extruded polygonal prisms. Basis functions are formed by the tensor product of 2D polygonal Bernstein–Bézier functions and 1D Lagrange polynomials. For a polynomial degree p, these functions span {xayb}(a+b)p{zc}c(0,p) with a dimension of np(p+1)+(p+1)(p1)(p2)/2 on an extruded n-gon. Numerical tests confirm that the functions capture exactly monomial solutions, achieve expected convergence rates, and provide full resolution in the thick diffusion limit. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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10 pages, 927 KiB  
Article
Self-Consistent Energy Normalization for Quasistatic Reactor Calculations
by David P. Griesheimer, Steven J. Douglass and Mark H. Stedry
J. Nucl. Eng. 2021, 2(2), 215-224; https://doi.org/10.3390/jne2020020 - 2 Jun 2021
Cited by 5 | Viewed by 2716
Abstract
Use of the quasistatic (keff) approximation for producing steady-state solutions for non-critical fission systems is known to result in an imbalance between energy release and deposition within the system. In this paper, we formally quantify this imbalance and present a [...] Read more.
Use of the quasistatic (keff) approximation for producing steady-state solutions for non-critical fission systems is known to result in an imbalance between energy release and deposition within the system. In this paper, we formally quantify this imbalance and present a self-consistent energy normalization technique that preserves nuclear energy release per reaction, as well as enforces energy balance between release and deposition mechanisms, regardless of the criticality state of the system. The proposed technique is straightforward to implement in any type of transport solver through the use of a simple energy rebalance factor. Theoretical and numerical results are presented that demonstrate the energy deposition bias for non-critical systems and the effectiveness of the proposed energy normalization technique. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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8 pages, 1710 KiB  
Article
Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER
by Thinh Truong, Heikki Suikkanen and Juhani Hyvärinen
J. Nucl. Eng. 2021, 2(2), 207-214; https://doi.org/10.3390/jne2020019 - 1 Jun 2021
Cited by 4 | Viewed by 4030
Abstract
In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel [...] Read more.
In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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7 pages, 1002 KiB  
Article
Using Generalized Basis for Functional Expansion
by Zhuoran Han, Benoit Forget and Kord Smith
J. Nucl. Eng. 2021, 2(2), 161-167; https://doi.org/10.3390/jne2020016 - 29 Apr 2021
Cited by 1 | Viewed by 2170
Abstract
Functional expansion has been rigorously studied as a promising method in stochastic neutron transport and multi-physics coupling. It is a method to represent data specified on a desired domain as an expansion of basis set in a continuous manner. For convenience, the basis [...] Read more.
Functional expansion has been rigorously studied as a promising method in stochastic neutron transport and multi-physics coupling. It is a method to represent data specified on a desired domain as an expansion of basis set in a continuous manner. For convenience, the basis set for functional expansion is typically chosen to be orthogonal. In cylindrical PWR pin-cell simulations, the orthogonal Zernike polynomials have been used. The main advantage of using functional expansion in nuclear modeling is that it requires less memory to represent temperature and nuclide variations in fuel then using a fine discretization. Fewer variables are involved in the data storage and transfer process. Each nuclide can have its unique expansion order, which becomes very important for depletion problems. In a recent study, performance analysis was conducted on Zernike-based FETs on a 2D PWR geometry. For reaction rates like the absorption rate of U-238, however, many orders are needed with Zernike-based FETs to achieve a reasonable accuracy. This gap inspires the study in this paper on alternative basis set that can better capture the steep gradient with fewer orders. In this paper, a generalized functional expansion method is established. The basis set can be an arbitrary series of independent functions. To capture the self-shielding effect of U-238 absorption rate, an exponential basis set is chosen. The results show that the expansion order utilizing exponential basis can reduce by half of that from using orthogonal Zernike polynomials while achieving the same accuracy. The integrated reaction rate is also demonstrated to be preserved. This paper also shows that the generalized functional expansion could be a heuristic method for further investigation on continuous depletion problems. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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9 pages, 11156 KiB  
Article
A Topology Optimization Procedure for Assisting the Design of Nuclear Components
by Sébastien Chabod
J. Nucl. Eng. 2021, 2(2), 152-160; https://doi.org/10.3390/jne2020015 - 29 Apr 2021
Cited by 1 | Viewed by 1750
Abstract
In this paper, we show that a module implemented in the MCNP transport code to perform sensitivity analyses can be diverted to perform topology optimizations of nuclear equipment. Component design with this approach leads to sophisticated solutions that outperform their human-designed counterparts. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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20 pages, 9123 KiB  
Article
Comparison of Direct and Adjoint k and α-Eigenfunctions
by Vito Vitali, Florent Chevallier, Alexis Jinaphanh, Andrea Zoia and Patrick Blaise
J. Nucl. Eng. 2021, 2(2), 132-151; https://doi.org/10.3390/jne2020014 - 19 Apr 2021
Cited by 1 | Viewed by 1942
Abstract
Modal expansions based on k-eigenvalues and α-eigenvalues are commonly used in order to investigate the reactor behaviour, each with a distinct point of view: the former is related to fission generations, whereas the latter is related to time. Well-known Monte Carlo [...] Read more.
Modal expansions based on k-eigenvalues and α-eigenvalues are commonly used in order to investigate the reactor behaviour, each with a distinct point of view: the former is related to fission generations, whereas the latter is related to time. Well-known Monte Carlo methods exist to compute the direct k or α fundamental eigenmodes, based on variants of the power iteration. The possibility of computing adjoint eigenfunctions in continuous-energy transport has been recently implemented and tested in the development version of TRIPOLI-4®, using a modified version of the Iterated Fission Probability (IFP) method for the adjoint α calculation. In this work we present a preliminary comparison of direct and adjoint k and α eigenmodes by Monte Carlo methods, for small deviations from criticality. When the reactor is exactly critical, i.e., for k0 = 1 or equivalently α0 = 0, the fundamental modes of both eigenfunction bases coincide, as expected on physical grounds. However, for non-critical systems the fundamental k and α eigenmodes show significant discrepancies. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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8 pages, 1618 KiB  
Article
Structure-Dependent Doppler Broadening Using a Generalized Thermal Scattering Law
by Nina C. Fleming and Ayman I. Hawari
J. Nucl. Eng. 2021, 2(2), 124-131; https://doi.org/10.3390/jne2020013 - 8 Apr 2021
Cited by 1 | Viewed by 2603
Abstract
The thermal scattering law (TSL), i.e., S(α,β), represents the momentum and energy exchange phase space for a material. The incoherent and coherent components of the TSL correlate an atom’s trajectory with itself and/or with other atoms in the lattice structure. This structural [...] Read more.
The thermal scattering law (TSL), i.e., S(α,β), represents the momentum and energy exchange phase space for a material. The incoherent and coherent components of the TSL correlate an atom’s trajectory with itself and/or with other atoms in the lattice structure. This structural information is especially important for low energies where the wavelength of neutrons is on the order of the lattice interatomic spacing. Both thermal neutron scattering as well as low energy resonance broadening involve processes where incoming neutron responses are lattice dependent. Traditionally, Doppler broadening for absorption resonances approximates these interactions by assuming a Maxwell–Boltzmann distribution for the neutron velocity. For high energies and high temperatures, this approximation is reasonable. However, for low temperatures or low energies, the lattice structure binding effects will influence the velocity distribution. Using the TSL to determine the Doppler broadening directly introduces the material structure into the calculation to most accurately capture the momentum and energy space. Typically, the TSL is derived assuming cubic lattice symmetry. This approximation collapses the directional lattice information, including the polarization vectors and associated energies, into an energy-dependent function called the density of states. The cubic approximation, while valid for highly symmetric and uniformly bonded materials, is insufficient to capture the true structure. In this work, generalized formulation for the exact, lattice-dependent TSL is implemented within the Full Law Analysis Scattering System Hub (FLASSH) using polarization vectors and associated energies as fundamental input. These capabilities are utilized to perform the generalized structure Doppler broadening analysis for UO2. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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10 pages, 301 KiB  
Article
Illustrating Important Effects of Second-Order Sensitivities on Response Uncertainties in Reactor Physics
by Dan G. Cacuci
J. Nucl. Eng. 2021, 2(2), 114-123; https://doi.org/10.3390/jne2020012 - 8 Apr 2021
Viewed by 1666
Abstract
This paper illustrates the relative importance of the largest first- and second-order sensitivities of the leakage response of an OECD/NEA reactor physics benchmark (a polyethylene-reflected plutonium sphere) to the benchmark’s underlying total cross sections. It will be shown that numerous 2nd-order sensitivities of [...] Read more.
This paper illustrates the relative importance of the largest first- and second-order sensitivities of the leakage response of an OECD/NEA reactor physics benchmark (a polyethylene-reflected plutonium sphere) to the benchmark’s underlying total cross sections. It will be shown that numerous 2nd-order sensitivities of the leakage response with respect to the total cross sections are significantly larger than the largest corresponding 1st-order sensitivities. In particular, the contributions of the 2nd-order sensitivities cause the mean (expected) value of the response to differ appreciably from its computed value and also cause the response distribution to be skewed towards positive values relative to the mean. Neglecting these large 2nd-order sensitivities would cause very large non-conservative errors by under-reporting the response’s variance and expected value. The results presented in this paper also underscore the need for obtaining reliable cross section covariance data, which are currently unavailable. Finally, comparing the CPU-times needed for computations, this paper demonstrates that the Second-Order Adjoint Sensitivity Analysis Methodology is the only practical method for computing 2nd-order sensitivities exactly, without introducing methodological errors, for large-scale systems characterized by many uncertain parameters. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
9 pages, 1348 KiB  
Article
Generation of the TSL for Zirconium Hydrides from Ab Initio Methods
by Jonathan Wormald, Michael Zerkle and Jesse Holmes
J. Nucl. Eng. 2021, 2(2), 105-113; https://doi.org/10.3390/jne2020011 - 7 Apr 2021
Cited by 7 | Viewed by 2837
Abstract
Zirconium hydride (ZrHx) is a moderator material used in TRIGA and other reactors that may exist in multiple phases with varying stoichiometry, which include the δ phase and the ϵ phase. Current ENDF/B-VIII.0 ZrHx thermal scattering law (TSL) evaluations do [...] Read more.
Zirconium hydride (ZrHx) is a moderator material used in TRIGA and other reactors that may exist in multiple phases with varying stoichiometry, which include the δ phase and the ϵ phase. Current ENDF/B-VIII.0 ZrHx thermal scattering law (TSL) evaluations do not distinguish between phases. These sub-libraries were generated with the LEAPR module of NJOY using historic phonon spectra derived from a central force model and assume incoherent elastic scattering for both bound hydrogen and zirconium, which neglects the effects of crystal structures important for scattering from zirconium bound in ZrHx. In this work, the TSLs for hydrogen and zirconium bound in δ-ZrHx and ϵ-ZrH2 were generated from phonon spectra derived from modern ab initio lattice dynamics methods and ab initio molecular dynamics. Subsequently, TSLs for hydrogen and zirconium in ZrHx and ZrH2 were generated using the Full Law Analysis Scattering System Hub (FLASSH) code. The built-in generalized coherent elastic routine was used to generate the previously neglected elastic contribution from zirconium for this material. The present TSLs provide both a re-evaluation of the current ZrH sub-libraries and expansion of the set of TSLs available for the examination of neutrons in systems with zirconium hydride, permitting explicit treatment of δ and ϵ phases. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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8 pages, 643 KiB  
Article
Choosing Transport Events for Initiating Splitting and Rouletting
by Evan S. Gonzalez and Gregory G. Davidson
J. Nucl. Eng. 2021, 2(2), 97-104; https://doi.org/10.3390/jne2020010 - 26 Mar 2021
Cited by 1 | Viewed by 2027
Abstract
A study was performed to determine which transport events should be used to initiate a weight window lookup to achieve the best variance reduction performance. A weight window lookup potentially triggers particle splitting (in important regions of phase space) or rouletting (in unimportant [...] Read more.
A study was performed to determine which transport events should be used to initiate a weight window lookup to achieve the best variance reduction performance. A weight window lookup potentially triggers particle splitting (in important regions of phase space) or rouletting (in unimportant regions), thereby optimizing computational effort. Potential initiating transport events include collisions (both pre- and post-collision), geometry surface crossings, traversing a mean-free path, and streaming across a weight window boundary. Permutations of these initiating events were tested on an urban model with background radiation sources and a spent fuel cask with a neutron dose mesh tally. Generally, all methods perform better with finer weight window meshes. Tracking on weight windows performs well for coarse weight window meshes, while a combination of splitting each mean-free path, geometric surface crossing, and before collisions performs well for fine weight window meshes. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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11 pages, 2634 KiB  
Article
The Effect of the Flux Separability Approximation on Multigroup Neutron Transport
by Adam G. Nelson, William Boyd and Paul K. Romano
J. Nucl. Eng. 2021, 2(1), 86-96; https://doi.org/10.3390/jne2010009 - 22 Mar 2021
Cited by 6 | Viewed by 2288
Abstract
The angular dependence of flux-weighted multigroup cross sections is commonly neglected when generating multigroup libraries. The error of this flux separability approximation is typically not isolated from other error sources due to a lack of availability of library generation and corresponding solvers that [...] Read more.
The angular dependence of flux-weighted multigroup cross sections is commonly neglected when generating multigroup libraries. The error of this flux separability approximation is typically not isolated from other error sources due to a lack of availability of library generation and corresponding solvers that cannot relax this approximation. These errors can now be isolated and quantified with the availability of a multigroup Monte Carlo transport and multigroup library-generation capability in the OpenMC Monte Carlo transport code. This work will discuss relevant details of the OpenMC implementation, provide an example case useful for detailing the type of errors one can expect from making the flux separability approximation, and end with more realistic problems which show the impact of the approximation and highlight how it can strongly arise from an energy-dependent resonance absorption effect. Since the angle-dependence is intrinsically linked to the energy group structure, these examples also show that relaxing the flux separability approximation with angle-dependent cross sections could be used to reduce either the fine-tuning required to set a multigroup energy structure for a specific reactor type or the number of energy groups required to obtain a desired level of accuracy for a given problem. This trade-off could increase the costs of generating multigroup cross sections, and has the potential to require more memory for storing the multigroup library during the transport calculations, but it can significantly reduce the computational time required since the runtime of a discrete ordinates or method of characteristics neutron transport solver scales roughly linearly with the number of groups. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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12 pages, 7341 KiB  
Article
Candidate Core Designs for the Transformational Challenge Reactor
by Brian J. Ade, Benjamin R. Betzler, Aaron J. Wysocki, Michael S. Greenwood, Phillip C. Chesser, Kurt A. Terrani, Prashant K. Jain, Joseph R. Burns, Briana D. Hiscox, Jordan D. Rader, Jesse J. W. Heineman, Florent Heidet, Aurelien Bergeron, James W. Sterbentz, Tommy V. Holschuh, Nicholas R. Brown and Robert F. Kile
J. Nucl. Eng. 2021, 2(1), 74-85; https://doi.org/10.3390/jne2010008 - 19 Mar 2021
Cited by 7 | Viewed by 3116
Abstract
Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and [...] Read more.
Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YHx) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YHx moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems are assessed and summarized herein. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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9 pages, 2612 KiB  
Article
Validation of the MORET 5 Monte Carlo Transport Code on Reactor Physics Experiments
by Nicolas Leclaire and Isabelle Duhamel
J. Nucl. Eng. 2021, 2(1), 65-73; https://doi.org/10.3390/jne2010007 - 19 Mar 2021
Cited by 1 | Viewed by 1785
Abstract
The MORET 5 code, which has been developed over more than 50 years at IRSN, has recently evolved, in its continuous energy version, from a criticality oriented code to a code also focused on reactor physics applications. Some developments such as the implementation [...] Read more.
The MORET 5 code, which has been developed over more than 50 years at IRSN, has recently evolved, in its continuous energy version, from a criticality oriented code to a code also focused on reactor physics applications. Some developments such as the implementation of kinetics parameters contribute to that evolution. The aim of the paper is to present the validation of the code for the keff multiplication factor used in criticality studies as well as for other parameters commonly used in reactor physics applications. Special attention will be paid on commission tests performed in the CABRI French Reactor (CABRI is a pool-type research reactor operated by CEA and located in the Cadarache site in southern France used to simulate a sudden and instantaneous increase in power, known as a power transient, typical of a reactivity-initiated accident (RIA).) and the IPEN/MB-01 LCT-077 benchmark. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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8 pages, 565 KiB  
Article
SCONE: A Student-Oriented Modifiable Monte Carlo Particle Transport Framework
by Mikolaj Adam Kowalski, Paul Cosgrove, Jakob Broman and Eugene Shwageraus
J. Nucl. Eng. 2021, 2(1), 57-64; https://doi.org/10.3390/jne2010006 - 8 Mar 2021
Cited by 7 | Viewed by 3134
Abstract
Over the last decade, the importance of the Monte Carlo as a neutron transport calculation method has greatly increased. This paper describes a Monte Carlo particle transport framework SCONE, which aims to provide with easy-to-learn environment for graduate students to learn about Monte [...] Read more.
Over the last decade, the importance of the Monte Carlo as a neutron transport calculation method has greatly increased. This paper describes a Monte Carlo particle transport framework SCONE, which aims to provide with easy-to-learn environment for graduate students to learn about Monte Carlo methods and explore new ideas. The paper lists the steps taken to enhance new user experience of SCONE and briefly discuses how the architecture supports its goals. The current version of the code is compared against Serpent and shown to provide with sufficient accuracy to be used for teaching and proof-of-concept applications. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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13 pages, 6158 KiB  
Article
Innovations in Multi-Physics Methods Development, Validation, and Uncertainty Quantification
by Maria Avramova, Agustin Abarca, Jason Hou and Kostadin Ivanov
J. Nucl. Eng. 2021, 2(1), 44-56; https://doi.org/10.3390/jne2010005 - 7 Mar 2021
Cited by 14 | Viewed by 3255
Abstract
This paper provides a review of current and upcoming innovations in development, validation, and uncertainty quantification of nuclear reactor multi-physics simulation methods. Multi-physics modelling and simulations (M&S) provide more accurate and realistic predictions of the nuclear reactors behavior including local safety parameters. Multi-physics [...] Read more.
This paper provides a review of current and upcoming innovations in development, validation, and uncertainty quantification of nuclear reactor multi-physics simulation methods. Multi-physics modelling and simulations (M&S) provide more accurate and realistic predictions of the nuclear reactors behavior including local safety parameters. Multi-physics M&S tools can be subdivided in two groups: traditional multi-physics M&S on assembly/channel spatial scale (currently used in industry and regulation), and novel high-fidelity multi-physics M&S on pin (sub-pin)/sub-channel spatial scale. The current trends in reactor design and safety analysis are towards further development, verification, and validation of multi-physics multi-scale M&S combined with uncertainty quantification and propagation. Approaches currently applied for validation of the traditional multi-physics M&S are summarized and illustrated using established Nuclear Energy Agency/Organization for Economic Cooperation and Development (NEA/OECD) multi-physics benchmarks. Novel high-fidelity multi-physics M&S allow for insights crucial to resolve industry challenge and high impact problems previously impossible with the traditional tools. Challenges in validation of novel multi-physics M&S are discussed along with the needs for developing validation benchmarks based on experimental data. Due to their complexity, the novel multi-physics codes are still computationally expensive for routine applications. This fact motivates the use of high-fidelity novel models and codes to inform the low-fidelity traditional models and codes, leading to improved traditional multi-physics M&S. The uncertainty quantification and propagation across different scales (multi-scale) and multi-physics phenomena are demonstrated using the OECD/NEA Light Water Reactor Uncertainty Analysis in Modelling benchmark framework. Finally, the increasing role of data science and analytics techniques in development and validation of multi-physics M&S is summarized. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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9 pages, 2110 KiB  
Article
Optimisation of AGR-Like FHR Fuel Assembly Using Multi-Objective Particle Swarm Algorithm
by Marat Margulis and Eugene Shwageraus
J. Nucl. Eng. 2021, 2(1), 35-43; https://doi.org/10.3390/jne2010004 - 18 Feb 2021
Cited by 3 | Viewed by 2292
Abstract
Utilising molten salt as coolant instead of carbon dioxide in traditional advanced gas-cooled reactors (AGRs) can potentially increase their core power density, simplify the safety case and shorten the time needed for the development of the fluoride-salt-cooled high-temperature reactor (FHR). However, the change [...] Read more.
Utilising molten salt as coolant instead of carbon dioxide in traditional advanced gas-cooled reactors (AGRs) can potentially increase their core power density, simplify the safety case and shorten the time needed for the development of the fluoride-salt-cooled high-temperature reactor (FHR). However, the change of coolant has a strong impact on the system behaviour. Therefore, a new type of fuel assembly is required. However, the design of a new assembly is affected by a wide range of parameters. Systematic search through all the potential configurations is prohibitively computationally expensive. In this work, a multi objective particle swarm optimisation (MOPSO) algorithm is utilised to identify the most attractive candidate configurations for the hybrid AGR-like FHR assembly. The first optimisation step targets basic design parameters such as radius and enrichment of the fuel pins, their number and arrangement. MOPSO is based on the concept of Pareto dominance, which is used to determine the flight direction of the simulated particles. The outcome of the optimisation process provides insight on families of possible solutions, which described by the Pareto front. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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7 pages, 4258 KiB  
Article
As-Built Simulation of the High Flux Isotope Reactor
by Benjamin R. Betzler, David Chandler, Thomas M. Evans, Gregory G. Davidson, Charles R. Daily, Stephen C. Wilson and Scott W. Mosher
J. Nucl. Eng. 2021, 2(1), 28-34; https://doi.org/10.3390/jne2010003 - 7 Feb 2021
Cited by 2 | Viewed by 2252
Abstract
The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) is an 85 MWt flux trap-type research reactor that supports key research missions, including isotope production, materials irradiation, and neutron scattering. The core consists of an inner and an outer fuel element containing [...] Read more.
The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) is an 85 MWt flux trap-type research reactor that supports key research missions, including isotope production, materials irradiation, and neutron scattering. The core consists of an inner and an outer fuel element containing 171 and 369 involute-shaped plates, respectively. The thin fuel plates consist of a U3O8-Al dispersion fuel (highly enriched), an aluminum-based filler, and aluminum cladding. The fuel meat thickness is varied across the width of the involute plate to reduce thermal flux peaks at the radial edges of the fuel elements. Some deviation from the designed fuel meat shaping is allowed during manufacturing. A homogeneity scan of each fuel plate checks for potential anomalies in the fuel distribution by scanning the surface of the plate and comparing the attenuation of the beam to calibration standards. While typical HFIR simulations use homogenized fuel regions, explicit models of the plates were developed under the Low-Enriched Uranium Conversion Program. These explicit models typically include one inner and one outer fuel plate with nominal fuel distributions, and then the plates are duplicated to fill the space of the corresponding fuel element. Therefore, data extracted from these simulations are limited to azimuthally averaged quantities. To determine the reactivity and physics impacts of an as-built outer fuel element and generate azimuthally dependent data in the element, 369 unique fuel plate models were generated and positioned. This model generates the three-dimensional (i.e., radial–axial–azimuthal) plate power profile, where the azimuthal profile is impacted by features within the adjacent control element region and beryllium reflector. For an as-built model of the outer fuel element, plate-specific homogeneity data, 235U loading, enrichment, and channel thickness measurements were translated into the model, yielding a much more varied azimuthal power profile encompassed by uncertainty factors in analyses. These models were run with the ORNL-TN and Shift Monte Carlo tools, and they contained upwards of 500,000 cells and 100,000 unique tallies. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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8 pages, 848 KiB  
Article
Robustness Study of Electro-Nuclear Scenario under Disruption
by Jiali Liang, Marc Ernoult, Xavier Doligez, Sylvain David, Léa Tillard and Nicolas Thiollière
J. Nucl. Eng. 2021, 2(1), 1-8; https://doi.org/10.3390/jne2010001 - 28 Jan 2021
Cited by 8 | Viewed by 1710
Abstract
As the future of nuclear power is uncertain, only choosing one development objective for the coming decades can be risky; while trying to achieve several possible objectives at the same time may lead to a deadlock due to contradiction among them. In this [...] Read more.
As the future of nuclear power is uncertain, only choosing one development objective for the coming decades can be risky; while trying to achieve several possible objectives at the same time may lead to a deadlock due to contradiction among them. In this work, we study a simple scenario to illustrate the newly developed method of robustness study, which considers possible change of objectives. Starting from the current French fleet, two objectives are considered regarding the possible political choices for the future of nuclear power: A. Complete substitution of Pressurized Water Reactors by Sodium-cooled Fast Reactors in 2180; B. Minimization of all potential nuclear wastes without SFR deployment in 2180. To study the robustness of strategies, the disruption of objective is considered: the objective to be pursued is possibly changed abruptly from A into B at unknown time. To minimize the consequence of such uncertainty, the first option is to identify a robust static strategy, which shows the best performance for both objectives A and B in the predisruption situation. The second option is to adapt a trajectory which pursues initially objective A, for objective B in case of the disruption. To identify and to analyze the adaptively robust strategies, outcomes of possible adaptations upon a given trajectory are compared with the robust static optimum. The temporality of adaptive robustness is analyzed by investigating different adaptation times. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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7 pages, 2045 KiB  
Article
Nuclear Data Uncertainty Propagation in Complex Fusion Geometries
by Bor Kos, Henrik Sjöstrand, Ivan A. Kodeli and JET Contributors
J. Nucl. Eng. 2020, 1(1), 63-69; https://doi.org/10.3390/jne1010006 - 2 Dec 2020
Cited by 1 | Viewed by 2420
Abstract
The ASUSD program package was designed to automate and simplify the process of deterministic nuclear data sensitivity and uncertainty quantification. The program package couples Denovo, a discrete ordinate 3D transport solver, as part of ADVANTG and SUSD3D, a deterministic first order perturbation theory [...] Read more.
The ASUSD program package was designed to automate and simplify the process of deterministic nuclear data sensitivity and uncertainty quantification. The program package couples Denovo, a discrete ordinate 3D transport solver, as part of ADVANTG and SUSD3D, a deterministic first order perturbation theory based Sensitivity/Uncertainty code, using several auxiliary programs used for input data preparation and post processing. Because of the automation employed in ASUSD, it is useful for Sensitivity/Uncertainty analysis of complex fusion geometries. In this paper, ASUSD was used to quantify uncertainties in the JET KN2 irradiation position. The results were compared to previously obtained probabilistic-based uncertainties determined using TALYS-based random nuclear data samples and MCNP in a Total Monte Carlo computation scheme. Results of the two approaches, deterministic and probabilistic, to nuclear data uncertainty propagation are compared and discussed. ASUSD was also used to perform preliminary Sensitivity/Uncertainty (S/U) analyses of three JET3-NEXP streaming benchmark experimental positions (A1, A4 and A7). Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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9 pages, 1293 KiB  
Article
Artificial Neural Networks as Surrogate Models for Uncertainty Quantification and Data Assimilation in 2-D/3-D Fuel Performance Studies
by Carlo Fiorina, Alessandro Scolaro, Daniel Siefman, Mathieu Hursin and Andreas Pautz
J. Nucl. Eng. 2020, 1(1), 54-62; https://doi.org/10.3390/jne1010005 - 10 Nov 2020
Cited by 2 | Viewed by 2189
Abstract
This paper preliminarily investigates the use of data-driven surrogates for fuel performance codes. The objective is to develop fast-running models that can be used in the frame of uncertainty quantification and data assimilation studies. In particular, data assimilation techniques based on Monte Carlo [...] Read more.
This paper preliminarily investigates the use of data-driven surrogates for fuel performance codes. The objective is to develop fast-running models that can be used in the frame of uncertainty quantification and data assimilation studies. In particular, data assimilation techniques based on Monte Carlo sampling often require running several thousand, or tens of thousands of calculations. In these cases, the computational requirements can quickly become prohibitive, notably for 2-D and 3-D codes. The paper analyses the capability of artificial neural networks to model the steady-state thermal-mechanics of the nuclear fuel, assuming given released fission gases, swelling, densification and creep. An optimized and trained neural network is then employed on a data assimilation case based on the end of the first ramp of the IFPE Instrumented Fuel Assemblies 432. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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8 pages, 1165 KiB  
Article
Preliminary Core Design Study of Small Supercritical Fast Reactor with Single-Pass Cooling
by Kyota Uchimura and Akifumi Yamaji
J. Nucl. Eng. 2020, 1(1), 46-53; https://doi.org/10.3390/jne1010004 - 7 Nov 2020
Cited by 3 | Viewed by 2659
Abstract
A supercritical water-cooled reactor (SCWR) adopts a once-through direct cycle, which is compatible with a small modular reactor class (SMR) plant system. The core is cooled by supercritical light water, which does not exhibit phase change, but undergoes large temperature and density changes. [...] Read more.
A supercritical water-cooled reactor (SCWR) adopts a once-through direct cycle, which is compatible with a small modular reactor class (SMR) plant system. The core is cooled by supercritical light water, which does not exhibit phase change, but undergoes large temperature and density changes. A super fast reactor (Super FR) is a fast reactor type concept of SCWR. Unlike other SCWR core concepts, it adopts the single coolant pass flow scheme, in which the coolant passes the core only once from the bottom to the top without any reverse flows or preheating stages. In the meantime, reducing the core size tends to increase the core power peaking and reduce criticality. Therefore, the key issues with the small Super FR core design is reducing the core power peaking and achieving high average core outlet temperature with the single coolant pass scheme. This study aims to highlight the design issues through conceptual core designs of SMR class Super FR. To evaluate the core characteristics, three-dimensional coupled core calculations are carried out. The proposed design with small fuel assemblies, which are equivalent to those of boiling water reactors, attains a high core average outlet temperature of about 500 °C, which is compatible to that of typical large SCWR core design. Full article
(This article belongs to the Special Issue Selected Papers from PHYSOR 2020)
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