Journal Publications Connor Moreno, Aaron Bader, Paul Wilson, "ParaStell: parametric modeling and neutronics support for stellarator fusion power plants", Frontiers in Nuclear Engineering, 3, pp. (2024-04-04) Abstract BibTeX The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell's future, as well as its implementation in machine-driven optimization. @article{moreno_parastell_2024, title = {{ParaStell}: parametric modeling and neutronics support for stellarator fusion power plants}, volume = {3}, issn = {2813-3412}, shorttitle = {{ParaStell}}, url = {https://www.frontiersin.org/articles/10.3389/fnuen.2024.1384788}, doi = {10.3389/fnuen.2024.1384788}, abstract = {The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell's future, as well as its implementation in machine-driven optimization.}, language = {English}, urldate = {2024-04-06}, journal = {Frontiers in Nuclear Engineering}, author = {Moreno, Connor and Bader, Aaron and Wilson, Paul}, month = apr, year = {2024}, note = {Publisher: Frontiers}, keywords = {FIRE UP, Fusion, Parametric modeling, Tritium breeding, magnet heating, radial build, stellarator}, } Jordan R. Stomps, Paul P. H. Wilson, Kenneth J. Dayman, "Contrastive Machine Learning with Gamma Spectroscopy Data Augmentations for Detecting Shielded Radiological Material Transfers", Mathematics, 12, pp. 2518 (2024/1) Abstract BibTeX Data analysis techniques can be powerful tools for rapidly analyzing data and extracting information that can be used in a latent space for categorizing observations between classes of data. Machine learning models that exploit learned data relationships can address a variety of nuclear nonproliferation challenges like the detection and tracking of shielded radiological material transfers. The high resource cost of manually labeling radiation spectra is a hindrance to the rapid analysis of data collected from persistent monitoring and to the adoption of supervised machine learning methods that require large volumes of curated training data. Instead, contrastive self-supervised learning on unlabeled spectra can enhance models that are built on limited labeled radiation datasets. This work demonstrates that contrastive machine learning is an effective technique for leveraging unlabeled data in detecting and characterizing nuclear material transfers demonstrated on radiation measurements collected at an Oak Ridge National Laboratory testbed, where sodium iodide detectors measure gamma radiation emitted by material transfers between the High Flux Isotope Reactor and the Radiochemical Engineering Development Center. Label-invariant data augmentations tailored for gamma radiation detection physics are used on unlabeled spectra to contrastively train an encoder, learning a complex, embedded state space with self-supervision. A linear classifier is then trained on a limited set of labeled data to distinguish transfer spectra between byproducts and tracked nuclear material using representations from the contrastively trained encoder. The optimized hyperparameter model achieves a balanced accuracy score of 80.30%. Any given model—that is, a trained encoder and classifier—shows preferential treatment for specific subclasses of transfer types. Regardless of the classifier complexity, a supervised classifier using contrastively trained representations achieves higher accuracy than using spectra when trained and tested on limited labeled data. @article{stomps_contrastive_2024, title = {Contrastive {Machine} {Learning} with {Gamma} {Spectroscopy} {Data} {Augmentations} for {Detecting} {Shielded} {Radiological} {Material} {Transfers}}, volume = {12}, copyright = {http://creativecommons.org/licenses/by/3.0/}, issn = {2227-7390}, url = {https://www.mdpi.com/2227-7390/12/16/2518}, doi = {10.3390/math12162518}, abstract = {Data analysis techniques can be powerful tools for rapidly analyzing data and extracting information that can be used in a latent space for categorizing observations between classes of data. Machine learning models that exploit learned data relationships can address a variety of nuclear nonproliferation challenges like the detection and tracking of shielded radiological material transfers. The high resource cost of manually labeling radiation spectra is a hindrance to the rapid analysis of data collected from persistent monitoring and to the adoption of supervised machine learning methods that require large volumes of curated training data. Instead, contrastive self-supervised learning on unlabeled spectra can enhance models that are built on limited labeled radiation datasets. This work demonstrates that contrastive machine learning is an effective technique for leveraging unlabeled data in detecting and characterizing nuclear material transfers demonstrated on radiation measurements collected at an Oak Ridge National Laboratory testbed, where sodium iodide detectors measure gamma radiation emitted by material transfers between the High Flux Isotope Reactor and the Radiochemical Engineering Development Center. Label-invariant data augmentations tailored for gamma radiation detection physics are used on unlabeled spectra to contrastively train an encoder, learning a complex, embedded state space with self-supervision. A linear classifier is then trained on a limited set of labeled data to distinguish transfer spectra between byproducts and tracked nuclear material using representations from the contrastively trained encoder. The optimized hyperparameter model achieves a balanced accuracy score of 80.30\%. Any given model—that is, a trained encoder and classifier—shows preferential treatment for specific subclasses of transfer types. Regardless of the classifier complexity, a supervised classifier using contrastively trained representations achieves higher accuracy than using spectra when trained and tested on limited labeled data.}, language = {en}, number = {16}, urldate = {2024-10-15}, journal = {Mathematics}, author = {Stomps, Jordan R. and Wilson, Paul P. H. and Dayman, Kenneth J.}, month = jan, year = {2024}, note = {Number: 16 Publisher: Multidisciplinary Digital Publishing Institute}, keywords = {contrastive learning, data analysis, gamma-ray spectroscopy, nuclear nonproliferation, radiation monitoring, semi-supervised machine learning}, pages = {2518}, } Nancy Granda Duarte, Irina I. Popova, Erik B. Iverson, Franz X. Gallmeier, Paul P. H. Wilson, "Shutdown Dose Rate with Cartesian Mesh for High-Energy Nuclear Systems", Nuclear Technology, 209, pp. 1747-1764 (2023-11-02) Abstract BibTeX In accelerator-driven systems, charged particles and high-energy neutrons can contribute to the production of nuclides that can persist long after the system has been shut down. These nuclides release photons that contribute to the biological dose. It is essential to quantify the biological dose as a function of time after shutdown to ensure safe working conditions for laborers during maintenance procedures. The shutdown dose rate (SDR) can be calculated with the Rigorous Two-Step (R2S) method, which includes a neutron and photon transport coupled with an activation calculation. For accelerator-driven systems, calculating SDR presents challenges related to the neutron cross-sectional data available for high-energy neutrons. A tally was implemented to collect isotope production data directly in a Monte Carlo N-Particle (MCNP) calculation. The output of this RNUCS tally is then used directly in an activation calculation, bypassing the need to use cross-section data with the neutron flux to obtain the isotope production and destruction data. A mesh-based RNUCS-R2S workflow has been developed based on this tally to calculate SDR in accelerator-driven systems. This workflow operates directly on computer-aided design geometry and supports using a meshed photon source. This workflow has been verified against a cell-based RNUCS-R2S workflow. A test problem with the essential characteristics of an accelerator-driven system was created to use in this analysis. The SDR results are within 40% of the cell-based RNUCS-R2S results. The workflow was also validated with the spallation neutron source system. Most detectors’ SDR results are within 50%, with a few detectors having a significantly lower SDR result than the experimental value. @article{granda_duarte_shutdown_2023, title = {Shutdown {Dose} {Rate} with {Cartesian} {Mesh} for {High}-{Energy} {Nuclear} {Systems}}, volume = {209}, issn = {0029-5450}, url = {https://doi.org/10.1080/00295450.2023.2205554}, doi = {10.1080/00295450.2023.2205554}, abstract = {In accelerator-driven systems, charged particles and high-energy neutrons can contribute to the production of nuclides that can persist long after the system has been shut down. These nuclides release photons that contribute to the biological dose. It is essential to quantify the biological dose as a function of time after shutdown to ensure safe working conditions for laborers during maintenance procedures. The shutdown dose rate (SDR) can be calculated with the Rigorous Two-Step (R2S) method, which includes a neutron and photon transport coupled with an activation calculation. For accelerator-driven systems, calculating SDR presents challenges related to the neutron cross-sectional data available for high-energy neutrons. A tally was implemented to collect isotope production data directly in a Monte Carlo N-Particle (MCNP) calculation. The output of this RNUCS tally is then used directly in an activation calculation, bypassing the need to use cross-section data with the neutron flux to obtain the isotope production and destruction data. A mesh-based RNUCS-R2S workflow has been developed based on this tally to calculate SDR in accelerator-driven systems. This workflow operates directly on computer-aided design geometry and supports using a meshed photon source. This workflow has been verified against a cell-based RNUCS-R2S workflow. A test problem with the essential characteristics of an accelerator-driven system was created to use in this analysis. The SDR results are within 40\% of the cell-based RNUCS-R2S results. The workflow was also validated with the spallation neutron source system. Most detectors’ SDR results are within 50\%, with a few detectors having a significantly lower SDR result than the experimental value.}, number = {11}, urldate = {2023-10-10}, journal = {Nuclear Technology}, author = {Granda Duarte, Nancy and Popova, Irina I. and Iverson, Erik B. and Gallmeier, Franz X. and Wilson, Paul P. H.}, month = nov, year = {2023}, note = {Publisher: Taylor \& Francis \_eprint: https://doi.org/10.1080/00295450.2023.2205554}, keywords = {R2S, Rigorous Two-Step method, high-energy nuclear systems, radiation transport}, pages = {1747--1764}, } Jordan R. Stomps, Paul P. H. Wilson, Kenneth J. Dayman, Michael J. Willis, James M. Ghawaly, et al, "SNM Radiation Signature Classification Using Different Semi-Supervised Machine Learning Models", Journal of Nuclear Engineering, 4, pp. 448-466 (2023/9) Abstract BibTeX The timely detection of special nuclear material (SNM) transfers between nuclear facilities is an important monitoring objective in nuclear nonproliferation. Persistent monitoring enabled by successful detection and characterization of radiological material movements could greatly enhance the nuclear nonproliferation mission in a range of applications. Supervised machine learning can be used to signal detections when material is present if a model is trained on sufficient volumes of labeled measurements. However, the nuclear monitoring data needed to train robust machine learning models can be costly to label since radiation spectra may require strict scrutiny for characterization. Therefore, this work investigates the application of semi-supervised learning to utilize both labeled and unlabeled data. As a demonstration experiment, radiation measurements from sodium iodide (NaI) detectors are provided by the Multi-Informatics for Nuclear Operating Scenarios (MINOS) venture at Oak Ridge National Laboratory (ORNL) as sample data. Anomalous measurements are identified using a method of statistical hypothesis testing. After background estimation, an energy-dependent spectroscopic analysis is used to characterize an anomaly based on its radiation signatures. In the absence of ground-truth information, a labeling heuristic provides data necessary for training and testing machine learning models. Supervised logistic regression serves as a baseline to compare three semi-supervised machine learning models: co-training, label propagation, and a convolutional neural network (CNN). In each case, the semi-supervised models outperform logistic regression, suggesting that unlabeled data can be valuable when training and demonstrating value in semi-supervised nonproliferation implementations. @article{stomps_snm_2023, title = {{SNM} {Radiation} {Signature} {Classification} {Using} {Different} {Semi}-{Supervised} {Machine} {Learning} {Models}}, volume = {4}, copyright = {http://creativecommons.org/licenses/by/3.0/}, issn = {2673-4362}, url = {https://www.mdpi.com/2673-4362/4/3/32}, doi = {10.3390/jne4030032}, abstract = {The timely detection of special nuclear material (SNM) transfers between nuclear facilities is an important monitoring objective in nuclear nonproliferation. Persistent monitoring enabled by successful detection and characterization of radiological material movements could greatly enhance the nuclear nonproliferation mission in a range of applications. Supervised machine learning can be used to signal detections when material is present if a model is trained on sufficient volumes of labeled measurements. However, the nuclear monitoring data needed to train robust machine learning models can be costly to label since radiation spectra may require strict scrutiny for characterization. Therefore, this work investigates the application of semi-supervised learning to utilize both labeled and unlabeled data. As a demonstration experiment, radiation measurements from sodium iodide (NaI) detectors are provided by the Multi-Informatics for Nuclear Operating Scenarios (MINOS) venture at Oak Ridge National Laboratory (ORNL) as sample data. Anomalous measurements are identified using a method of statistical hypothesis testing. After background estimation, an energy-dependent spectroscopic analysis is used to characterize an anomaly based on its radiation signatures. In the absence of ground-truth information, a labeling heuristic provides data necessary for training and testing machine learning models. Supervised logistic regression serves as a baseline to compare three semi-supervised machine learning models: co-training, label propagation, and a convolutional neural network (CNN). In each case, the semi-supervised models outperform logistic regression, suggesting that unlabeled data can be valuable when training and demonstrating value in semi-supervised nonproliferation implementations.}, language = {en}, number = {3}, urldate = {2023-07-12}, journal = {Journal of Nuclear Engineering}, author = {Stomps, Jordan R. and Wilson, Paul P. H. and Dayman, Kenneth J. and Willis, Michael J. and Ghawaly, James M. and Archer, Daniel E.}, month = sep, year = {2023}, note = {Number: 3 Publisher: Multidisciplinary Digital Publishing Institute}, keywords = {data analysis, gamma-ray spectroscopy, nuclear nonproliferation, radiation monitoring, semi-supervised machine learning}, pages = {448--466}, } Joshua Ruegsegger, Connor Moreno, Matthew Nyberg, Tim Bohm, Paul P. H. Wilson, et al, "Scoping Studies for a Lead-Lithium-Cooled, Minor-Actinide-Burning, Fission-Fusion Hybrid Reactor Design", Nuclear Science and Engineering, , pp. 1-17 (2023-01-24) Abstract BibTeX A feasibility study of a subcritical fission-fusion hybrid reactor using lead-lithium eutectic as a coolant and minor actinides (MAs) as fuel is presented. Such a reactor could support the fission community by transmuting MAs and the fusion community by breeding tritium. The feasibility of such a reactor for the burnup of MAs is assessed in terms of burnup performance, tritium breeding, and safety characteristics. Tandem mirrors are a promising neutron source technology, and a deuterium-tritium tandem mirror is considered here for the neutron source with power Psource = 1.13 MW assumed for scoping purposes. Subcritical reactivities from keff = 0.9800 to keff = 0.9950 were considered, representing the initial reference for subcritical reactivity and the assessed upper limit, respectively. Stability analyses indicated the reactivity would be stable under perturbations of fuel, coolant, and inlet temperatures, with a positive reactivity insertion expected during reactor shutdown. This keff range corresponded to nuclear heating values of 150 to 650 MW and mass burn rates of 53 to 216 kg/year. The upper mass burn rate limit would require 1110 reactor years with a capacity factor of 0.9 to fission the global supply of MAs and could offset the annual U.S. MA production with eight reactors. Tritium breeding was assessed for keff = 0.9800 and 3.795% 6Li enrichment in the coolant, and a tritium breeding ratio of 1.602 ± 0.017 was tallied, suggesting the reactor could, without elevated 6Li enrichment, produce tritium to both sustain operation and supply tritium for other fusion devices. Time-series modeling of fuel burnup was conducted for a four-batch loading scheme and three different fuel residence times at keff = 0.9800, which showed system performance would drop with burnup, and that the rate of this drop was lower for longer fuel residence times, motivating a means of reactivity control. Last, changes in fuel composition with burnup were assessed for relative concentrations of MAs, transmutation products, and fission products. The breeding of plutonium in the blanket was calculated and found to be of minimal proliferation concern. @article{ruegsegger_scoping_2023, title = {Scoping {Studies} for a {Lead}-{Lithium}-{Cooled}, {Minor}-{Actinide}-{Burning}, {Fission}-{Fusion} {Hybrid} {Reactor} {Design}}, issn = {0029-5639}, url = {https://www.tandfonline.com/doi/abs/10.1080/00295639.2022.2154118}, doi = {10.1080/00295639.2022.2154118}, abstract = {A feasibility study of a subcritical fission-fusion hybrid reactor using lead-lithium eutectic as a coolant and minor actinides (MAs) as fuel is presented. Such a reactor could support the fission community by transmuting MAs and the fusion community by breeding tritium. The feasibility of such a reactor for the burnup of MAs is assessed in terms of burnup performance, tritium breeding, and safety characteristics. Tandem mirrors are a promising neutron source technology, and a deuterium-tritium tandem mirror is considered here for the neutron source with power Psource = 1.13 MW assumed for scoping purposes. Subcritical reactivities from keff = 0.9800 to keff = 0.9950 were considered, representing the initial reference for subcritical reactivity and the assessed upper limit, respectively. Stability analyses indicated the reactivity would be stable under perturbations of fuel, coolant, and inlet temperatures, with a positive reactivity insertion expected during reactor shutdown. This keff range corresponded to nuclear heating values of 150 to 650 MW and mass burn rates of 53 to 216 kg/year. The upper mass burn rate limit would require 1110 reactor years with a capacity factor of 0.9 to fission the global supply of MAs and could offset the annual U.S. MA production with eight reactors. Tritium breeding was assessed for keff = 0.9800 and 3.795\% 6Li enrichment in the coolant, and a tritium breeding ratio of 1.602 ± 0.017 was tallied, suggesting the reactor could, without elevated 6Li enrichment, produce tritium to both sustain operation and supply tritium for other fusion devices. Time-series modeling of fuel burnup was conducted for a four-batch loading scheme and three different fuel residence times at keff = 0.9800, which showed system performance would drop with burnup, and that the rate of this drop was lower for longer fuel residence times, motivating a means of reactivity control. Last, changes in fuel composition with burnup were assessed for relative concentrations of MAs, transmutation products, and fission products. The breeding of plutonium in the blanket was calculated and found to be of minimal proliferation concern.}, urldate = {2023-01-24}, journal = {Nuclear Science and Engineering}, author = {Ruegsegger, Joshua and Moreno, Connor and Nyberg, Matthew and Bohm, Tim and Wilson, Paul P. H. and Lindley, Ben}, month = jan, year = {2023}, note = {Publisher: Taylor \& Francis \_eprint: https://www.tandfonline.com/doi/pdf/10.1080/00295639.2022.2154118}, keywords = {Subcritical, burnup, fission-fusion hybrid, lead lithium, minor actinide}, pages = {1--17}, } M. Harb, A. Davis, P. P. H. Wilson, "Uncertainty Quantification of the Decay Gamma Source in Mesh-Based Shutdown Dose Rate Calculations", Fusion Science and Technology, 79, pp. 1-12 (2023-01-02) Abstract BibTeX In fusion energy systems, part of the design effort is dedicated to the assessment of the shutdown dose rate (SDR) due to the decay photons that will be emitted from activated components. Monte Carlo transport codes are often used to obtain the neutron flux distribution in the problem domain. The neutron flux distribution is used in the rigorous 2-step (R2S) workflow to obtain the photon emission density distribution of decaying radionuclides. The photon emission density is then used as an input for a dedicated photon transport step to calculate the SDR. In this paper, the uncertainty of the decay gamma source due to the uncertainty of the neutron flux distribution in the R2S workflow is investigated. A scheme is developed to estimate the uncertainty of the decay gamma source, building on the concept of groupwise transmutation and using standard error propagation techniques. The applicability of the newly developed scheme is then demonstrated on one of the conceptual designs of the fusion nuclear science facility. @article{harb_uncertainty_2023, title = {Uncertainty {Quantification} of the {Decay} {Gamma} {Source} in {Mesh}-{Based} {Shutdown} {Dose} {Rate} {Calculations}}, volume = {79}, issn = {1536-1055}, url = {https://doi.org/10.1080/15361055.2022.2115831}, doi = {10.1080/15361055.2022.2115831}, abstract = {In fusion energy systems, part of the design effort is dedicated to the assessment of the shutdown dose rate (SDR) due to the decay photons that will be emitted from activated components. Monte Carlo transport codes are often used to obtain the neutron flux distribution in the problem domain. The neutron flux distribution is used in the rigorous 2-step (R2S) workflow to obtain the photon emission density distribution of decaying radionuclides. The photon emission density is then used as an input for a dedicated photon transport step to calculate the SDR. In this paper, the uncertainty of the decay gamma source due to the uncertainty of the neutron flux distribution in the R2S workflow is investigated. A scheme is developed to estimate the uncertainty of the decay gamma source, building on the concept of groupwise transmutation and using standard error propagation techniques. The applicability of the newly developed scheme is then demonstrated on one of the conceptual designs of the fusion nuclear science facility.}, number = {1}, urldate = {2022-12-14}, journal = {Fusion Science and Technology}, author = {Harb, M. and Davis, A. and Wilson, P. P. H.}, month = jan, year = {2023}, note = {Publisher: Taylor \& Francis \_eprint: https://doi.org/10.1080/15361055.2022.2115831}, keywords = {DAGMC, FIRE MIT, Uncertainty propagation, fusion nuclear science facility, rigorous 2-step workflow, shutdown dose rate}, pages = {1--12}, } Xiaokang ZHANG, Patrick C. SHRIWISE, Songlin LIU, Paul P. H. WILSON, "Implementation and application of PyNE sub-voxel R2S for shutdown dose rate analysis", Plasma Science and Technology, 24, pp. 095603 (2022-07) Abstract BibTeX PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S. @article{zhang_implementation_2022, title = {Implementation and application of {PyNE} sub-voxel {R2S} for shutdown dose rate analysis}, volume = {24}, issn = {1009-0630}, url = {https://dx.doi.org/10.1088/2058-6272/ac6be3}, doi = {10.1088/2058-6272/ac6be3}, abstract = {PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S.}, language = {en}, number = {9}, urldate = {2022-10-27}, journal = {Plasma Science and Technology}, author = {ZHANG, Xiaokang and SHRIWISE, Patrick C. and LIU, Songlin and WILSON, Paul P. H.}, month = jul, year = {2022}, note = {Publisher: IOP Publishing}, pages = {095603}, } N. Thiollière, X. Doligez, M. Halasz, G. Krivtchik, I. Merino, et al, "Impact of fresh fuel loading management in fuel cycle simulators: A functionality isolation test", Nuclear Engineering and Design, 392, pp. 111748 (2022-06-01) Abstract BibTeX Fuel cycle simulator development started many years ago by several research and engineering institutions or consulting firms for a wide range of applications. To improve confidence in the results, institutions may be tempted to increase the complexity of their software even if this complexity might not be necessary. On the other hand, some simulators may be used outside their range of validity when used in very specific applications. The FIT (Functionality Isolation Test) project is an international effort devoted to improve the confidence in the data produced by fuel cycle simulation tools. The scientific goal is to determine the optimum level of detail a fuel cycle simulator needs according to the type of study and the required confidence level. The project relies on a wide variety of fuel cycle simulators with a large range of complexity levels. The FIT project consists of isolating the impact of one targeted functionality on fuel cycle simulations. The impact of the functionality is assessed using a set of simple basic exercises specifically designed for this purpose, called ”functionality isolation.” The present work focuses on the impact on simulation results of using a fuel loading model (a relation that links the stock isotopic composition with the fresh fuel fabrication according to the reactor requirements) or a fixed fraction approach (the fresh fuel fissile fraction is fixed and does not depend on the stock isotopic composition). The paper first presents the FIT project. The exercise design is described and results show that using a fuel loading model approach has an important impact on fuel cycle outputs under certain conditions that are described. This result is reinforced by the fact that all fuel cycle simulators used in this exercise provide similar conclusions. @article{thiolliere_impact_2022, title = {Impact of fresh fuel loading management in fuel cycle simulators: {A} functionality isolation test}, volume = {392}, issn = {0029-5493}, shorttitle = {Impact of fresh fuel loading management in fuel cycle simulators}, url = {https://www.sciencedirect.com/science/article/pii/S0029549322001029}, doi = {10.1016/j.nucengdes.2022.111748}, abstract = {Fuel cycle simulator development started many years ago by several research and engineering institutions or consulting firms for a wide range of applications. To improve confidence in the results, institutions may be tempted to increase the complexity of their software even if this complexity might not be necessary. On the other hand, some simulators may be used outside their range of validity when used in very specific applications. The FIT (Functionality Isolation Test) project is an international effort devoted to improve the confidence in the data produced by fuel cycle simulation tools. The scientific goal is to determine the optimum level of detail a fuel cycle simulator needs according to the type of study and the required confidence level. The project relies on a wide variety of fuel cycle simulators with a large range of complexity levels. The FIT project consists of isolating the impact of one targeted functionality on fuel cycle simulations. The impact of the functionality is assessed using a set of simple basic exercises specifically designed for this purpose, called ”functionality isolation.” The present work focuses on the impact on simulation results of using a fuel loading model (a relation that links the stock isotopic composition with the fresh fuel fabrication according to the reactor requirements) or a fixed fraction approach (the fresh fuel fissile fraction is fixed and does not depend on the stock isotopic composition). The paper first presents the FIT project. The exercise design is described and results show that using a fuel loading model approach has an important impact on fuel cycle outputs under certain conditions that are described. This result is reinforced by the fact that all fuel cycle simulators used in this exercise provide similar conclusions.}, language = {en}, urldate = {2022-04-18}, journal = {Nuclear Engineering and Design}, author = {Thiollière, N. and Doligez, X. and Halasz, M. and Krivtchik, G. and Merino, I. and Mouginot, B. and Skarbeli, A. V. and Hernandez-Solis, A. and Alvarez-Velarde, F. and Courtin, F. and Druenne, H. and Ernoult, M. and Huff, K. and Szieberth, M. and Vermeeren, B. and Wilson, P.}, month = jun, year = {2022}, keywords = {FIT project, Fuel Cycle Simulators, Fuel Loading Models, NEWTON, Pressurized Water Reactors, Sodium Fast Reactors}, pages = {111748}, } Patrick C. Shriwise, Paul P. H. Wilson, Andrew Davis, Paul K. Romano, "Hardware-Accelerated Ray Tracing of CAD-Based Geometry for Monte Carlo Radiation Transport", Computing in Science & Engineering, 24, pp. 52-61 (2022-03) Abstract BibTeX Monte Carlo radiation transport (MCRT) methods have been used to simulate radiation environments for many decades by tracking individual particles through a model to accumulate statistical information. MCRT geometry is historically formed using the constructive solid geometry (CSG). Recently, significant work has been performed to support simulations using computer-aided design (CAD)-based tessellated surfaces to support highly complex geometries. Ray tracing acceleration data structures from the rendering and visualization community are applied to accelerate particle tracking in CAD-based models. Despite these efforts, CSG representations provide the superior performance in surface intersection operations during particle flight. Concurrently, pseudo Monte Carlo methods have become prevalent in rendering applications to support more realistic models for scattering media, motivating innovations that are advantageous for MCRT simulations. The authors’ work extends these innovations by employing Intel’s Embree ray tracing kernel within a geometry toolkit for Monte Carlo to improve the simulation performance using CAD-based models by factors of 1.5 to 2. @article{shriwise_hardware-accelerated_2022, title = {Hardware-{Accelerated} {Ray} {Tracing} of {CAD}-{Based} {Geometry} for {Monte} {Carlo} {Radiation} {Transport}}, volume = {24}, issn = {1558-366X}, url = {https://ieeexplore.ieee.org/document/9721666}, doi = {10.1109/MCSE.2022.3154656}, abstract = {Monte Carlo radiation transport (MCRT) methods have been used to simulate radiation environments for many decades by tracking individual particles through a model to accumulate statistical information. MCRT geometry is historically formed using the constructive solid geometry (CSG). Recently, significant work has been performed to support simulations using computer-aided design (CAD)-based tessellated surfaces to support highly complex geometries. Ray tracing acceleration data structures from the rendering and visualization community are applied to accelerate particle tracking in CAD-based models. Despite these efforts, CSG representations provide the superior performance in surface intersection operations during particle flight. Concurrently, pseudo Monte Carlo methods have become prevalent in rendering applications to support more realistic models for scattering media, motivating innovations that are advantageous for MCRT simulations. The authors’ work extends these innovations by employing Intel’s Embree ray tracing kernel within a geometry toolkit for Monte Carlo to improve the simulation performance using CAD-based models by factors of 1.5 to 2.}, number = {2}, journal = {Computing in Science \& Engineering}, author = {Shriwise, Patrick C. and Wilson, Paul P. H. and Davis, Andrew and Romano, Paul K.}, month = mar, year = {2022}, note = {Conference Name: Computing in Science \& Engineering}, keywords = {Computational modeling, Geometry, Hardware acceleration, Mathematical models, Monte Carlo methods, Ray tracing, Solid modeling}, pages = {52--61}, } Stephen Buono, Jake Hecla, Vladimir Kobezskii, Katie Mummah, Julien de Troullioud de Lanversin, "It’s time to reignite US-Russia cooperation in space. Nuclear power may hold the key", Bulletin of the Atomic Scientists, 77, pp. 203-206 (July 4, 2021) Abstract BibTeX As US-Russia tensions in space have increased over the last several years, cooperation in space nuclear research presents itself as one opportunity to both ease bilateral relations and develop the technologies needed for the next generation of crewed space missions. The authors suggest that the United States and the Russian Federation have complementary needs and strengths in nuclear space technologies, particularly as they pertain to deep space propulsion and utilization of space resources. @article{buono_its_2021, title = {It’s time to reignite {US}-{Russia} cooperation in space. {Nuclear} power may hold the key}, volume = {77}, issn = {0096-3402}, url = {https://doi.org/10.1080/00963402.2021.1941598}, doi = {10.1080/00963402.2021.1941598}, abstract = {As US-Russia tensions in space have increased over the last several years, cooperation in space nuclear research presents itself as one opportunity to both ease bilateral relations and develop the technologies needed for the next generation of crewed space missions. The authors suggest that the United States and the Russian Federation have complementary needs and strengths in nuclear space technologies, particularly as they pertain to deep space propulsion and utilization of space resources.}, number = {4}, urldate = {2022-02-24}, journal = {Bulletin of the Atomic Scientists}, author = {Buono, Stephen and Hecla, Jake and Kobezskii, Vladimir and Mummah, Katie and de Troullioud de Lanversin, Julien}, month = jul, year = {2021}, note = {Publisher: Routledge \_eprint: https://doi.org/10.1080/00963402.2021.1941598}, keywords = {Mars Oxygen In-Situ Resource Utilization Experiment, TOPAZ International Program, US-Russia space cooperation, in situ resource utilization, nuclear thermal propulsion}, pages = {203--206}, } Young-Hui Park, Ye Cheng, Rabab Elzohery, Paul P. H. Wilson, Jeremy A. Roberts, et al, "EVALUATION OF CRITICAL EXPERIMENTS IN THE UNIVERSITY OF WISCONSIN NUCLEAR REACTOR (UWNR) WITH UNCERTAINTY QUANTIFICATION", EPJ Web of Conferences, 247, pp. 10032 (2021) Abstract BibTeX An improved computational model of the University of Wisconsin Nuclear Reactor (UWNR) was developed to support the benchmark evaluation of recent data acquired during an experimental campaign conducted at UWNR. Previous efforts led to a scripted UWNR model for automated generation of MCNP6 and Serpent inputs. This capability was extended to SCALE/KENO. All three tools were used to evaluate a variety of zero-power, fresh-critical configurations, and the results agreed well. The MCNP6 model was extended to support shuffling the core configuration, which allows the modeling of burnup for evaluation of depleted critical configurations. The MCNP6 model successfully predicts core reactivity over time, after accounting for the initial reactivity bias. The inclusion of SCALE/KENO input generation enables sensitivity and uncertainty analyses using the TSUNAMI and Sampler modules of SCALE. A preliminary uncertainty analysis was performed with TSUNAMI for nuclear data uncertainties while direct perturbation calculations were performed using MCNP6 for geometry and material uncertainties, which helped to identify model parameters with the largest effect on the eigenvalue. A transient UWNR transport Model in Mammoth/Rattlesnake is under development to simulate the transient experiments. The existing MCNP6 and Serpent models are used to provide the CAD file for meshing and homogenized cross-sections. In conclusion, the evaluation of UWNR benchmark data provides increased confidence in various states of the UWNR computational model and will provide a unique model for use by other analysts. @article{park_evaluation_2021, title = {{EVALUATION} {OF} {CRITICAL} {EXPERIMENTS} {IN} {THE} {UNIVERSITY} {OF} {WISCONSIN} {NUCLEAR} {REACTOR} ({UWNR}) {WITH} {UNCERTAINTY} {QUANTIFICATION}}, volume = {247}, copyright = {© The Authors, published by EDP Sciences, 2021}, issn = {2100-014X}, url = {https://www.epj-conferences.org/articles/epjconf/abs/2021/01/epjconf_physor2020_10032/epjconf_physor2020_10032.html}, doi = {10.1051/epjconf/202124710032}, abstract = {An improved computational model of the University of Wisconsin Nuclear Reactor (UWNR) was developed to support the benchmark evaluation of recent data acquired during an experimental campaign conducted at UWNR. Previous efforts led to a scripted UWNR model for automated generation of MCNP6 and Serpent inputs. This capability was extended to SCALE/KENO. All three tools were used to evaluate a variety of zero-power, fresh-critical configurations, and the results agreed well. The MCNP6 model was extended to support shuffling the core configuration, which allows the modeling of burnup for evaluation of depleted critical configurations. The MCNP6 model successfully predicts core reactivity over time, after accounting for the initial reactivity bias. The inclusion of SCALE/KENO input generation enables sensitivity and uncertainty analyses using the TSUNAMI and Sampler modules of SCALE. A preliminary uncertainty analysis was performed with TSUNAMI for nuclear data uncertainties while direct perturbation calculations were performed using MCNP6 for geometry and material uncertainties, which helped to identify model parameters with the largest effect on the eigenvalue. A transient UWNR transport Model in Mammoth/Rattlesnake is under development to simulate the transient experiments. The existing MCNP6 and Serpent models are used to provide the CAD file for meshing and homogenized cross-sections. In conclusion, the evaluation of UWNR benchmark data provides increased confidence in various states of the UWNR computational model and will provide a unique model for use by other analysts.}, language = {en}, urldate = {2022-08-16}, journal = {EPJ Web of Conferences}, author = {Park, Young-Hui and Cheng, Ye and Elzohery, Rabab and Wilson, Paul P. H. and Roberts, Jeremy A. and DeHart, Mark D.}, year = {2021}, note = {Publisher: EDP Sciences}, pages = {10032}, } C. E. Kessel, D. Andruczyk, J. P. Blanchard, T. Bohm, A. Davis, et al, "Critical Exploration of Liquid Metal Plasma-Facing Components in a Fusion Nuclear Science Facility", Fusion Science and Technology, 75, pp. 886-917 (November 17, 2019) Abstract BibTeX Liquid metal (LM) plasma-facing components (PFCs) may provide a resolution to the challenging fusion environment, particularly the first wall and divertor surfaces. Transforming these concepts into viable technologies will require considerable research and development. With the fusion nuclear regime in mind, the Fusion Energy System Studies group examined LM PFCs in order to identify needed research thrusts that could accelerate their development and assess their viability. Liquid metal behavior, solid substrate aspects, and fusion facility integration aspects are examined, with concepts as the research focusing element. The concepts applied to a fusion nuclear device are the primary definer of the LM parameters, environmental conditions, and operational aspects. This forms the research strategy recommended for these complex systems. @article{kessel_critical_2019, title = {Critical {Exploration} of {Liquid} {Metal} {Plasma}-{Facing} {Components} in a {Fusion} {Nuclear} {Science} {Facility}}, volume = {75}, issn = {1536-1055}, url = {https://doi.org/10.1080/15361055.2019.1610685}, doi = {10.1080/15361055.2019.1610685}, abstract = {Liquid metal (LM) plasma-facing components (PFCs) may provide a resolution to the challenging fusion environment, particularly the first wall and divertor surfaces. Transforming these concepts into viable technologies will require considerable research and development. With the fusion nuclear regime in mind, the Fusion Energy System Studies group examined LM PFCs in order to identify needed research thrusts that could accelerate their development and assess their viability. Liquid metal behavior, solid substrate aspects, and fusion facility integration aspects are examined, with concepts as the research focusing element. The concepts applied to a fusion nuclear device are the primary definer of the LM parameters, environmental conditions, and operational aspects. This forms the research strategy recommended for these complex systems.}, number = {8}, urldate = {2020-05-01}, journal = {Fusion Science and Technology}, author = {Kessel, C. E. and Andruczyk, D. and Blanchard, J. P. and Bohm, T. and Davis, A. and Hollis, K. and Humrickhouse, P. W. and Hvasta, M. and Jaworski, M. and Jun, J. and Katoh, Y. and Khodak, A. and Klein, J. and Kolemen, E. and Larsen, G. and Majeski, R. and Merrill, B. J. and Morley, N. B. and Neilson, G. H. and Pint, B. and Rensink, M. E. and Rognlien, T. D. and Rowcliffe, A. F. and Smolentsev, S. and Tillack, M. S. and Waganer, L. M. and Wallace, G. M. and Wilson, P. and Yoon, S.-J.}, month = nov, year = {2019}, keywords = {Fusion Nuclear Science Facility, Liquid metals, design, plasma-facing components, tokamak}, pages = {886--917}, } M. Harb, T. Bohm, A. Davis, P. P. H. Wilson, the FESS-FNSF Team, "Calculation of Shutdown Dose Rate in Fusion Nuclear Science Facility During a Proposed Maintenance Scheme", Fusion Science and Technology, 75, pp. 747-753 (October 3, 2019) Abstract BibTeX In this work, a preliminary assessment of the shutdown dose rate (SDR) in the latest Fusion Energy Systems Studies–Fusion Nuclear Science Facility conceptual design was calculated for one sector at different maintenance stages. The third operational phase, deuterium-tritium for 2.75 years, was considered to define the neutron source and the Rigorous 2-Step workflow was used. SDR levels were obtained at times that correspond to major maintenance operations and were found to be above 105 µSv/h, which necessitates robotic handling of all maintenance operations. @article{harb_calculation_2019, title = {Calculation of {Shutdown} {Dose} {Rate} in {Fusion} {Nuclear} {Science} {Facility} {During} a {Proposed} {Maintenance} {Scheme}}, volume = {75}, issn = {1536-1055}, url = {https://doi.org/10.1080/15361055.2019.1644134}, doi = {10.1080/15361055.2019.1644134}, abstract = {In this work, a preliminary assessment of the shutdown dose rate (SDR) in the latest Fusion Energy Systems Studies–Fusion Nuclear Science Facility conceptual design was calculated for one sector at different maintenance stages. The third operational phase, deuterium-tritium for 2.75 years, was considered to define the neutron source and the Rigorous 2-Step workflow was used. SDR levels were obtained at times that correspond to major maintenance operations and were found to be above 105 µSv/h, which necessitates robotic handling of all maintenance operations.}, number = {7}, urldate = {2019-10-09}, journal = {Fusion Science and Technology}, author = {Harb, M. and Bohm, T. and Davis, A. and Wilson, P. P. H. and Team, the FESS-FNSF}, month = oct, year = {2019}, keywords = {FESS-FNSF, Rigorous 2-Step, DAGMC, shutdown dose rate}, pages = {747--753}, } M. Harb, T. Bohm, A. Davis, P. P. H. Wilson, , "Calculation of Shutdown Dose Rate in Fusion Nuclear Science Facility During a Proposed Maintenance Scheme", Fusion Science and Technology, 75, pp. 747-753 (2019-10-03) BibTeX @article{harb_calculation_2019, title = {Calculation of {Shutdown} {Dose} {Rate} in {Fusion} {Nuclear} {Science} {Facility} {During} a {Proposed} {Maintenance} {Scheme}}, volume = {75}, issn = {1536-1055, 1943-7641}, url = {https://www.tandfonline.com/doi/full/10.1080/15361055.2019.1644134}, doi = {10.1080/15361055.2019.1644134}, language = {en}, number = {7}, urldate = {2024-03-28}, journal = {Fusion Science and Technology}, author = {Harb, M. and Bohm, T. and Davis, A. and Wilson, P. P. H. and {the FESS-FNSF Team}}, month = oct, year = {2019}, pages = {747--753}, } E. Joffrin, S. Abduallev, M. Abhangi, P. Abreu, V. Afanasev, et al, "Overview of the JET preparation for deuterium–tritium operation with the ITER like-wall", Nuclear Fusion, 59, pp. 112021 (August 2019) Abstract BibTeX For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfvèn eigenmode antennas, neutral gauges, radiation hard imaging systems…) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas. @article{joffrin_overview_2019, title = {Overview of the {JET} preparation for deuterium–tritium operation with the {ITER} like-wall}, volume = {59}, issn = {0029-5515}, url = {https://doi.org/10.1088%2F1741-4326%2Fab2276}, doi = {10.1088/1741-4326/ab2276}, abstract = {For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50\%/50\% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfvèn eigenmode antennas, neutral gauges, radiation hard imaging systems…) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.}, language = {en}, number = {11}, urldate = {2020-05-01}, journal = {Nuclear Fusion}, author = {Joffrin, E. and Abduallev, S. and Abhangi, M. and Abreu, P. and Afanasev, V. and Afzal, M. and Aggarwal, K. M. and Ahlgren, T. and Aho-Mantila, L. and Aiba, N. and Airila, M. and Alarcon, T. and Albanese, R. and Alegre, D. and Aleiferis, S. and Alessi, E. and Aleynikov, P. and Alkseev, A. and Allinson, M. and Alper, B. and Alves, E. and Ambrosino, G. and Ambrosino, R. and Amosov, V. and Sundén, E. Andersson and Andrews, R. and Angelone, M. and Anghel, M. and Angioni, C. and Appel, L. and Appelbee, C. and Arena, P. and Ariola, M. and Arshad, S. and Artaud, J. and Arter, W. and Ash, A. and Ashikawa, N. and Aslanyan, V. and Asunta, O. and Asztalos, O. and Auriemma, F. and Austin, Y. and Avotina, L. and Axton, M. and Ayres, C. and Baciero, A. and Baião, D. and Balboa, I. and Balden, M. and Balshaw, N. and Bandaru, V. K. and Banks, J. and Baranov, Y. F. and Barcellona, C. and Barnard, T. and Barnes, M. and Barnsley, R. and Wiechec, A. Baron and Orte, L. Barrera and Baruzzo, M. and Basiuk, V. and Bassan, M. and Bastow, R. and Batista, A. and Batistoni, P. and Baumane, L. and Bauvir, B. and Baylor, L. and Beaumont, P. S. and Beckers, M. and Beckett, B. and Bekris, N. and Beldishevski, M. and Bell, K. and Belli, F. and Belonohy, É and Benayas, J. and Bergs{\textbackslash}aaker, H. and Bernardo, J. and Bernert, M. and Berry, M. and Bertalot, L. and Besiliu, C. and Betar, H. and Beurskens, M. and Bielecki, J. and Biewer, T. and Bilato, R. and Biletskyi, O. and Bílková, P. and Binda, F. and Birkenmeier, G. and Bizarro, J. P. S. and Björkas, C. and Blackburn, J. and Blackman, T. R. and Blanchard, P. and Blatchford, P. and Bobkov, V. and Boboc, A. and Bogar, O. and Bohm, P. and Bohm, T. and Bolshakova, I. and Bolzonella, T. and Bonanomi, N. and Boncagni, L. and Bonfiglio, D. and Bonnin, X. and Boom, J. and Borba, D. and Borodin, D. and Borodkina, I. and Boulbe, C. and Bourdelle, C. and Bowden, M. and Bowman, C. and Boyce, T. and Boyer, H. and Bradnam, S. C. and Braic, V. and Bravanec, R. and Breizman, B. and Brennan, D. and Breton, S. and Brett, A. and Brezinsek, S. and Bright, M. and Brix, M. and Broeckx, W. and Brombin, M. and Bros{\textbackslash}lawski, A. and Brown, B. and Brunetti, D. and Bruno, E. and Buch, J. and Buchanan, J. and Buckingham, R. and Buckley, M. and Bucolo, M. and Budny, R. and Bufferand, H. and Buller, S. and Bunting, P. and Buratti, P. and Burckhart, A. and Burroughes, G. and Buscarino, A. and Busse, A. and Butcher, D. and Butler, B. and Bykov, I. and Cahyna, P. and Calabrò, G. and Calacci, L. and Callaghan, D. and Callaghan, J. and Calvo, I. and Camenen, Y. and Camp, P. and Campling, D. C. and Cannas, B. and Capat, A. and Carcangiu, S. and Card, P. and Cardinali, A. and Carman, P. and Carnevale, D. and Carr, M. and Carralero, D. and Carraro, L. and Carvalho, B. B. and Carvalho, I. and Carvalho, P. and Carvalho, D. D. and Casson, F. J. and Castaldo, C. and Catarino, N. and Causa, F. and Cavazzana, R. and Cave-Ayland, K. and Cavedon, M. and Cecconello, M. and Ceccuzzi, S. and Cecil, E. and Challis, C. D. and Chandra, D. and Chang, C. S. and Chankin, A. and Chapman, I. T. and Chapman, B. and Chapman, S. C. and Chernyshova, M. and Chiariello, A. and Chitarin, G. and Chmielewski, P. and Chone, L. and Ciraolo, G. and Ciric, D. and Citrin, J. and Clairet, F. and Clark, M. and Clark, E. and Clarkson, R. and Clay, R. and Clements, C. and Coad, J. P. and Coates, P. and Cobalt, A. and Coccorese, V. and Cocilovo, V. and Coelho, R. and Coenen, J. W. and Coffey, I. and Colas, L. and Colling, B. and Collins, S. and Conka, D. and Conroy, S. and Conway, N. and Coombs, D. and Cooper, S. R. and Corradino, C. and Corre, Y. and Corrigan, G. and Coster, D. and Craciunescu, T. and Cramp, S. and Crapper, C. and Crisanti, F. and Croci, G. and Croft, D. and Crombé, K. and Cruz, N. and Cseh, G. and Cufar, A. and Cullen, A. and Curson, P. and Curuia, M. and Czarnecka, A. and Czarski, T. and Cziegler, I. and Dabirikhah, H. and Molin, A. Dal and Dalgliesh, P. and Dalley, S. and Dankowski, J. and Darrow, D. and David, P. and Davies, A. and Davis, W. and Dawson, K. and Day, I. and Day, C. and Bock, M. De and Castro, A. de and Dominici, G. De and Cal, E. de la and Luna, E. de la and Masi, G. De and Temmerman, G. De and Tommasi, G. De and Vries, P. de and Deane, J. and Dejarnac, R. and Sarto, D. Del and Delabie, E. and Demerdzhiev, V. and Dempsey, A. and Harder, N. den and Dendy, R. O. and Denis, J. and Denner, P. and Devaux, S. and Devynck, P. and Maio, F. Di and Siena, A. Di and Troia, C. Di and Dickinson, D. and Dinca, P. and Dittmar, T. and Dobrashian, J. and Doerk, H. and Doerner, R. P. and Domptail, F. and Donné, T. and Dorling, S. E. and Douai, D. and Dowson, S. and Drenik, A. and Dreval, M. and Drewelow, P. and Drews, P. and Duckworth, Ph and Dumont, R. and Dumortier, P. and Dunai, D. and Dunne, M. and Ďuran, I. and Durodié, F. and Dutta, P. and Duval, B. P. and Dux, R. and Dylst, K. and Edappala, P. V. and Edwards, A. M. and Edwards, J. S. and Eich, Th and Eidietis, N. and Eksaeva, A. and Ellis, R. and Ellwood, G. and Elsmore, C. and Emery, S. and Enachescu, M. and Ericsson, G. and Eriksson, J. and Eriksson, F. and Eriksson, L. G. and Ertmer, S. and Esquembri, S. and Esquisabel, A. L. and Esser, H. G. and Ewart, G. and Fable, E. and Fagan, D. and Faitsch, M. and Falie, D. and Fanni, A. and Farahani, A. and Fasoli, A. and Faugeras, B. and Fazinic{\textbackslash}', S. and Felici, F. and Felton, R. C. and Feng, S. and Fernades, A. and Fernandes, H. and Ferreira, J. and Ferreira, D. R. and Ferrò, G. and Fessey, J. A. and Ficker, O. and Field, A. and Fietz, S. and Figini, L. and Figueiredo, J. and Figueiredo, A. and Figueiredo, J. and Fil, N. and Finburg, P. and Fischer, U. and Fittill, L. and Fitzgerald, M. and Flammini, D. and Flanagan, J. and Flinders, K. and Foley, S. and Fonnesu, N. and Fontdecaba, J. 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R. and Wischmeier, M. and Withycombe, A. and Witts, D. and Wolfrum, E. and Wood, R. and Woodley, R. and Woodley, C. and Wray, S. and Wright, J. C. and Wright, P. and Wukitch, S. and Wynn, A. and Xiang, L. and Xu, T. and Xue, Y. and Yadikin, D. and Yakovenko, Y. and Yanling, W. and Yavorskij, V. and Young, I. and Young, R. and Young, D. and Zacks, J. and Zagorski, R. and Zaitsev, F. S. and Zakharov, L. and Zanino, R. and Zarins, A. and Zarins, R. and Fernandez, D. Zarzoso and Zastrow, K. D. and Zerbini, M. and Zhang, W. and Zhou, Y. and Zilli, E. and Zocco, A. and Zoita, V. and Zoletnik, S. and Zwingmann, W. and Zychor, I.}, month = aug, year = {2019}, pages = {112021}, } Tim D. Bohm, Andrew Davis, Moataz S. Harb, Edward P. Marriott, Paul P. H. Wilson, "Initial Neutronics Investigation of a Liquid-Metal Plasma-Facing Fusion Nuclear Science Facility", Fusion Science and Technology, 75, pp. 429-437 (May 6, 2019) Abstract BibTeX The use of a liquid-metal (LM) plasma-facing component (LM-PFC) in fusion reactor designs has some advantages as well as some disadvantages as compared to traditional designs that use a solid plasma-facing wall. Neutronics analysis of these potential LM-PFC concepts is important in order to ensure that radiation limits are met and that system performance meets expectations.A three-dimensional (3-D) neutronics analysis parametric study considering four LM first-wall (FW) candidates, (PbLi, Li, Sn, and SnLi) was performed with a thin (2.51-cm) LM-PFC design. The 3-D neutronics study used a fusion reactor based on the Fusion Energy Systems Study (FESS) Fusion Nuclear Science Facility (FNSF) (FESS-FNSF) that served as the baseline for comparison. FESS-FNSF is a deuterium-tritium–fueled tokamak with 518 MW of fusion power. A partially homogenized 3-D computer-aided-design model of the LM-PFC FNSF design was analyzed using the DAG-MCNP5 transport code.The results show that all candidate LM designs are acceptable with 4% to 13% increases in the tritium breeding ratio compared to the baseline case. The peak displacements per atom at the FW decrease 2% to 15%. For all four LM designs examined, the magnet heating and fast neutron fluence are well below acceptable limits. Overall, the Li LM design is the best candidate from a neutronics perspective. @article{bohm_initial_2019, title = {Initial {Neutronics} {Investigation} of a {Liquid}-{Metal} {Plasma}-{Facing} {Fusion} {Nuclear} {Science} {Facility}}, volume = {75}, issn = {1536-1055}, url = {https://doi.org/10.1080/15361055.2019.1600930}, doi = {10.1080/15361055.2019.1600930}, abstract = {The use of a liquid-metal (LM) plasma-facing component (LM-PFC) in fusion reactor designs has some advantages as well as some disadvantages as compared to traditional designs that use a solid plasma-facing wall. Neutronics analysis of these potential LM-PFC concepts is important in order to ensure that radiation limits are met and that system performance meets expectations.A three-dimensional (3-D) neutronics analysis parametric study considering four LM first-wall (FW) candidates, (PbLi, Li, Sn, and SnLi) was performed with a thin (2.51-cm) LM-PFC design. The 3-D neutronics study used a fusion reactor based on the Fusion Energy Systems Study (FESS) Fusion Nuclear Science Facility (FNSF) (FESS-FNSF) that served as the baseline for comparison. FESS-FNSF is a deuterium-tritium–fueled tokamak with 518 MW of fusion power. A partially homogenized 3-D computer-aided-design model of the LM-PFC FNSF design was analyzed using the DAG-MCNP5 transport code.The results show that all candidate LM designs are acceptable with 4\% to 13\% increases in the tritium breeding ratio compared to the baseline case. The peak displacements per atom at the FW decrease 2\% to 15\%. For all four LM designs examined, the magnet heating and fast neutron fluence are well below acceptable limits. Overall, the Li LM design is the best candidate from a neutronics perspective.}, number = {6}, urldate = {2019-05-22}, journal = {Fusion Science and Technology}, author = {Bohm, Tim D. and Davis, Andrew and Harb, Moataz S. and Marriott, Edward P. and Wilson, Paul P. H.}, month = may, year = {2019}, keywords = {DAG-MCNP, Fusion Nuclear Science Facility, nuclear heating, radiation damage, tritium breeding ratio}, pages = {429--437}, } Andrew Davis, Janet Barzilla, Alfredo Ferrari, Kerry T. Lee, Vasillis Vlachoudis, et al, "FluDAG: A CAD based tool for high energy physics", Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 915, pp. 65-74 (January 21, 2019) Abstract BibTeX As part of the expansion of the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit to support other Monte Carlo codes, FluDAG (FLUKA integrated with DAGMC) was developed. There has been increasing demand from the high energy physics community regarding Computer Aided Design (CAD) geometry support in Monte Carlo codes. In this paper, the development and validation of FluDAG is discussed and its application to a number of high energy physics experiments is demonstrated, along with its validity relative to native FLUKA calculations. @article{davis_fludag:_2019, title = {{FluDAG}: {A} {CAD} based tool for high energy physics}, volume = {915}, issn = {0168-9002}, shorttitle = {{FluDAG}}, url = {http://www.sciencedirect.com/science/article/pii/S0168900218314311}, doi = {10.1016/j.nima.2018.10.117}, abstract = {As part of the expansion of the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit to support other Monte Carlo codes, FluDAG (FLUKA integrated with DAGMC) was developed. There has been increasing demand from the high energy physics community regarding Computer Aided Design (CAD) geometry support in Monte Carlo codes. In this paper, the development and validation of FluDAG is discussed and its application to a number of high energy physics experiments is demonstrated, along with its validity relative to native FLUKA calculations.}, urldate = {2019-03-21}, journal = {Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment}, author = {Davis, Andrew and Barzilla, Janet and Ferrari, Alfredo and Lee, Kerry T. and Vlachoudis, Vasillis and Wilson, Paul P. H.}, month = jan, year = {2019}, keywords = {CAD, DAGMC, FLUKA, Monte Carlo, Radiation physics}, pages = {65--74}, } Jeff Radtke, Paul Deluca, Dan Anderson, Paul Wilson, Laura Bartol, et al, "Ionization Chambers to Determine Neutron and Gamma-Ray Kerma in a Research Reactor", IEEE Transactions on Nuclear Science, 66, pp. 2160-2169 (2019) Abstract BibTeX Ionization chambers were designed and constructed to determine the kerma rates in various materials within several centimeters of a TRIGA reactor core operating at 1 MW. The primary aim of this work was to compare kerma measurements with advanced Monte Carlo code calculations of nuclear heating. Wall thickness, collection gap and fill gas pressure were chosen to satisfy Bragg-Gray criteria, so that measured ionization current was related to the kerma rate in the wall material. Chamber wall materials comprised of low mass number elements, including hydrogen-rich C552 air-equivalent plastic and beryllium, were selected to measure the kerma due to fast neutron elastic scattering. By operating these neutron sensitive chambers coincidentally with relatively neutron insensitive chambers composed of aluminum and Zircaloy-4, we were able to measure the total heating due to fast neutrons and gamma rays in a material, and differentiate these heating components. A chamber comprised of borated stainless steel was used in a similar fashion to measure thermal neutron flux. The total kerma rate was also measured in various materials typically found in a reactor core. Chamber collection volumes were initially determined using ambient air fill gas and NIST-traceable air-kerma rates from a 60Co source. All chambers were sealed with argon gas to provide thermal and compositional stability. Chamber properties, including stability, saturation and gas phase mass subject to charge collection, were determined using the 60Co source. Chambers were operated for approximately 30 minutes adjacent the reactor core, and integrity of gas seals was subsequently verified by repeating the measurement with the 60Co source. @article{radtke_ionization_2019, title = {Ionization {Chambers} to {Determine} {Neutron} and {Gamma}-{Ray} {Kerma} in a {Research} {Reactor}}, volume = {66}, issn = {0018-9499, 1558-1578}, url = {https://ieeexplore.ieee.org/document/8815849}, doi = {10.1109/TNS.2019.2937750}, abstract = {Ionization chambers were designed and constructed to determine the kerma rates in various materials within several centimeters of a TRIGA reactor core operating at 1 MW. The primary aim of this work was to compare kerma measurements with advanced Monte Carlo code calculations of nuclear heating. Wall thickness, collection gap and fill gas pressure were chosen to satisfy Bragg-Gray criteria, so that measured ionization current was related to the kerma rate in the wall material. Chamber wall materials comprised of low mass number elements, including hydrogen-rich C552 air-equivalent plastic and beryllium, were selected to measure the kerma due to fast neutron elastic scattering. By operating these neutron sensitive chambers coincidentally with relatively neutron insensitive chambers composed of aluminum and Zircaloy-4, we were able to measure the total heating due to fast neutrons and gamma rays in a material, and differentiate these heating components. A chamber comprised of borated stainless steel was used in a similar fashion to measure thermal neutron flux. The total kerma rate was also measured in various materials typically found in a reactor core. Chamber collection volumes were initially determined using ambient air fill gas and NIST-traceable air-kerma rates from a 60Co source. All chambers were sealed with argon gas to provide thermal and compositional stability. Chamber properties, including stability, saturation and gas phase mass subject to charge collection, were determined using the 60Co source. Chambers were operated for approximately 30 minutes adjacent the reactor core, and integrity of gas seals was subsequently verified by repeating the measurement with the 60Co source.}, journal = {IEEE Transactions on Nuclear Science}, author = {Radtke, Jeff and Deluca, Paul and Anderson, Dan and Wilson, Paul and Bartol, Laura and Maile, Andrew and Agasie, Robert and Trumbull, Timothy and Grant, Edwin and Brooks, Paul and Anderson, Mark and Culberson, Wesley}, year = {2019}, keywords = {Bragg-Gray, Kerma, TRIGA, ionization chamber, neutron-gamma discrimination, nuclear heating}, pages = {2160--2169}, } Kalin Kiesling, Paul P. H. Wilson, "Generation of Weight Window Isosurface Geometries for Monte-Carlo Variance Reduction", Transactions of the American Nuclear Society, 119, pp. 586-589 (11/13/2018) BibTeX @article{kiesling_generation_2018, title = {Generation of {Weight} {Window} {Isosurface} {Geometries} for {Monte}-{Carlo} {Variance} {Reduction}}, volume = {119}, journal = {Transactions of the American Nuclear Society}, author = {Kiesling, Kalin and Wilson, Paul P. H.}, month = nov, year = {2018}, note = {Orlando, FL November 11-15, 2018}, keywords = {DAGMC, Weight window, isosurface}, pages = {586--589}, } Patrick Shriwise, Paul Wilson, "Reduced Precision Ray Tracing Performance Enhancements in DAGMC", Transactions of the American Nuclear Society, 119, pp. 608-611 (11/13/2018) BibTeX @article{shriwise_reduced_2018, title = {Reduced {Precision} {Ray} {Tracing} {Performance} {Enhancements} in {DAGMC}}, volume = {119}, journal = {Transactions of the American Nuclear Society}, author = {Shriwise, Patrick and Wilson, Paul}, month = nov, year = {2018}, pages = {608--611}, } Chelsea D'Angelo, Paul P. H. Wilson, "Generating Variance Reduction Parameters for Shutdown Dose Rate Analysis of Moving Systems", Transactions of the American Nuclear Society, 119, pp. (November, 2018) BibTeX @article{dangelo_generating_2018, title = {Generating {Variance} {Reduction} {Parameters} for {Shutdown} {Dose} {Rate} {Analysis} of {Moving} {Systems}}, volume = {119}, journal = {Transactions of the American Nuclear Society}, author = {D'Angelo, Chelsea and Wilson, Paul P. H.}, month = nov, year = {2018}, } A. Davis, M. Harb, L. El-Guebaly, P. Wilson, E. Marriott, "Neutronics aspects of the FESS-FNSF", Fusion Engineering and Design, 135, pp. 271-278 (October 1, 2018) Abstract BibTeX Neutronics analysis was performed on the latest Fusion Energy System Studies-Fusion Nuclear Science Facility (FESS-FNSF) design, which determined the neutron wall loading, tritium breeding ratio, and radiation damage. Sixteen different sectors configurations were investigated, with the main focus on determining the impact which each has upon the tritium breeding ratio (TBR) of the whole facility. This paper describes the stages of the nuclear analysis that serve to prove the radiation derived attributes of the system. @article{davis_neutronics_2018, series = {Special {Issue}: {FESS}-{FNSF} {Study}}, title = {Neutronics aspects of the {FESS}-{FNSF}}, volume = {135}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S0920379617307081}, doi = {10.1016/j.fusengdes.2017.06.008}, abstract = {Neutronics analysis was performed on the latest Fusion Energy System Studies-Fusion Nuclear Science Facility (FESS-FNSF) design, which determined the neutron wall loading, tritium breeding ratio, and radiation damage. Sixteen different sectors configurations were investigated, with the main focus on determining the impact which each has upon the tritium breeding ratio (TBR) of the whole facility. This paper describes the stages of the nuclear analysis that serve to prove the radiation derived attributes of the system.}, urldate = {2018-09-20}, journal = {Fusion Engineering and Design}, author = {Davis, A. and Harb, M. and El-Guebaly, L. and Wilson, P. and Marriott, E.}, month = oct, year = {2018}, keywords = {DAGMC, FESS-FNSF, Radiation damage, Tritium breeding ratio}, pages = {271--278}, } Nan Li, Dominique Brossard, Dietram A. Scheufele, Paul H. Wilson, Kathleen M. Rose, "Communicating data: interactive infographics, scientific data and credibility", Journal of Science Communication, 17, pp. A06 (2018/06/18) Abstract BibTeX Information visualization could be used to leverage the credibility of displayed scientific data. However, little was known about how display characteristics interact with individuals' predispositions to affect perception of data credibility. Using an experiment with 517 participants, we tested perceptions of data credibility by manipulating data visualizations related to the issue of nuclear fuel cycle based on three characteristics: graph format, graph interactivity, and source attribution. Results showed that viewers tend to rely on preexisting levels of trust and peripheral cues, such as source attribution, to judge the credibility of shown data, whereas their comprehension level did not relate to perception of data credibility. We discussed the implications for science communicators and design professionals. @article{li_communicating_2018, title = {Communicating data: interactive infographics, scientific data and credibility}, volume = {17}, issn = {1824-2049}, shorttitle = {Communicating data}, url = {https://jcom.sissa.it/archive/17/02/JCOM_1702_2018_A06}, doi = {10.22323/2.17020206}, abstract = {Information visualization could be used to leverage the credibility of displayed scientific data. However, little was known about how display characteristics interact with individuals' predispositions to affect perception of data credibility. Using an experiment with 517 participants, we tested perceptions of data credibility by manipulating data visualizations related to the issue of nuclear fuel cycle based on three characteristics: graph format, graph interactivity, and source attribution. Results showed that viewers tend to rely on preexisting levels of trust and peripheral cues, such as source attribution, to judge the credibility of shown data, whereas their comprehension level did not relate to perception of data credibility. We discussed the implications for science communicators and design professionals.}, language = {en}, number = {2}, urldate = {2019-03-21}, journal = {Journal of Science Communication}, author = {Li, Nan and Brossard, Dominique and Scheufele, Dietram A. and Wilson, Paul H. and Rose, Kathleen M.}, month = jun, year = {2018}, keywords = {NEWTON}, pages = {A06}, } Nan Li, Dominique Brossard, Dietram A. Scheufele, Paul P. H. Wilson, "Policymakers and stakeholders' perceptions of science-driven nuclear energy policy", Nuclear Engineering and Technology, 50, pp. 773-779 (June 1, 2018) Abstract BibTeX This study surveyed 137 policymakers and key stakeholders (e.g., employees of government agencies, academic institutions, nonprofit organizations, industry, and advocacy groups) involved in making decisions on nuclear energy policy, investigating how they differentially perceived the importance of scientific evidence in driving nuclear policy. We also identified the policy areas that each group of decision-makers are mostly concerned about and showed how such concerns might contextualize and ultimately shape their perceptions of science-driven policy. @article{li_policymakers_2018, title = {Policymakers and stakeholders' perceptions of science-driven nuclear energy policy}, volume = {50}, issn = {1738-5733}, url = {http://www.sciencedirect.com/science/article/pii/S1738573318300068}, doi = {10.1016/j.net.2018.03.012}, abstract = {This study surveyed 137 policymakers and key stakeholders (e.g., employees of government agencies, academic institutions, nonprofit organizations, industry, and advocacy groups) involved in making decisions on nuclear energy policy, investigating how they differentially perceived the importance of scientific evidence in driving nuclear policy. We also identified the policy areas that each group of decision-makers are mostly concerned about and showed how such concerns might contextualize and ultimately shape their perceptions of science-driven policy.}, number = {5}, urldate = {2018-06-14}, journal = {Nuclear Engineering and Technology}, author = {Li, Nan and Brossard, Dominique and Scheufele, Dietram A. and Wilson, Paul P. H.}, month = jun, year = {2018}, keywords = {NEWTON, Nuclear Energy, Nuclear Fuel Cycles, Science-driven Policy, Scientist–Policymaker Communication}, pages = {773--779}, } M. Harb, L. El-Guebaly, A. Davis, P. Wilson, E. Marriott, et al, "3-D Neutronics Assessment of Tritium Breeding Capacity and Shielding of Tokamak-Based Fusion Nuclear Science Facility", Fusion Science and Technology, 72, pp. 510-515 (October 3, 2017) Abstract BibTeX Two issues related to neutronics analysis of fusion systems were addressed for the purpose of physical design iterations as well as plant operation: tritium self-sufficiency and shielding of the inboard magnet. State-of-the-art modeling/analysis tools facilitated a full 3-D neutronics analysis of the latest FESS-FNSF design. The first stage of the analysis involved the selection of materials for the first wall and blanket along with shielding materials to protect the magnet based on extensive 1-D analyses. The second stage is a stepwise workflow to estimate the overall tritium breeding ratio with high fidelity. It involved a bottom-up approach by coupling the CAD model with the 3-D MCNP code using DAGMC and adding the relevant design details in steps to assess the effect of such details on the tritium breeding ratio. The final stage involved calculations of the values of damage parameters at specific components: the first wall, the vacuum vessel, and magnet. @article{harb_3-d_2017, title = {3-{D} {Neutronics} {Assessment} of {Tritium} {Breeding} {Capacity} and {Shielding} of {Tokamak}-{Based} {Fusion} {Nuclear} {Science} {Facility}}, volume = {72}, issn = {1536-1055}, url = {https://doi.org/10.1080/15361055.2017.1333846}, doi = {10.1080/15361055.2017.1333846}, abstract = {Two issues related to neutronics analysis of fusion systems were addressed for the purpose of physical design iterations as well as plant operation: tritium self-sufficiency and shielding of the inboard magnet. State-of-the-art modeling/analysis tools facilitated a full 3-D neutronics analysis of the latest FESS-FNSF design. The first stage of the analysis involved the selection of materials for the first wall and blanket along with shielding materials to protect the magnet based on extensive 1-D analyses. The second stage is a stepwise workflow to estimate the overall tritium breeding ratio with high fidelity. It involved a bottom-up approach by coupling the CAD model with the 3-D MCNP code using DAGMC and adding the relevant design details in steps to assess the effect of such details on the tritium breeding ratio. The final stage involved calculations of the values of damage parameters at specific components: the first wall, the vacuum vessel, and magnet.}, number = {3}, urldate = {2019-10-09}, journal = {Fusion Science and Technology}, author = {Harb, M. and El-Guebaly, L. and Davis, A. and Wilson, P. and Marriott, E. and Benzaquen, J. and Team, FESS-FNSF}, month = oct, year = {2017}, keywords = {DAGMC, FESS-FNSF, radiation damage, tritium breeding ratio}, pages = {510--515}, } Patrick C. Shriwise, Andrew Davis, Lucas J. Jacobson, Paul P. H. Wilson, "Particle tracking acceleration via signed distance fields in direct-accelerated geometry Monte Carlo", Nuclear Engineering and Technology, 49, pp. 1189-1198 (September 1, 2017) Abstract BibTeX Computer-aided design (CAD)-based Monte Carlo radiation transport is of value to the nuclear engineering community for its ability to conduct transport on high-fidelity models of nuclear systems, but it is more computationally expensive than native geometry representations. This work describes the adaptation of a rendering data structure, the signed distance field, as a geometric query tool for accelerating CAD-based transport in the direct-accelerated geometry Monte Carlo toolkit. Demonstrations of its effectiveness are shown for several problems. The beginnings of a predictive model for the data structure's utilization based on various problem parameters is also introduced. @article{shriwise_particle_2017, series = {Special {Issue} on {International} {Conference} on {Mathematics} and {Computational} {Methods} {Applied} to {Nuclear} {Science} and {Engineering} 2017 ({M}\&{C} 2017)}, title = {Particle tracking acceleration via signed distance fields in direct-accelerated geometry {Monte} {Carlo}}, volume = {49}, issn = {1738-5733}, url = {http://www.sciencedirect.com/science/article/pii/S1738573317303145}, doi = {10.1016/j.net.2017.08.008}, abstract = {Computer-aided design (CAD)-based Monte Carlo radiation transport is of value to the nuclear engineering community for its ability to conduct transport on high-fidelity models of nuclear systems, but it is more computationally expensive than native geometry representations. This work describes the adaptation of a rendering data structure, the signed distance field, as a geometric query tool for accelerating CAD-based transport in the direct-accelerated geometry Monte Carlo toolkit. Demonstrations of its effectiveness are shown for several problems. The beginnings of a predictive model for the data structure's utilization based on various problem parameters is also introduced.}, number = {6}, urldate = {2017-11-24}, journal = {Nuclear Engineering and Technology}, author = {Shriwise, Patrick C. and Davis, Andrew and Jacobson, Lucas J. and Wilson, Paul P. H.}, month = sep, year = {2017}, keywords = {CAD, DAGMC, Monte Carlo, Radiation Transport, product}, pages = {1189--1198}, } L. El-Guebaly, L. Mynsberge, A. Davis, C. D’Angelo, A. Rowcliffe, et al, "Design and Evaluation of Nuclear System for ARIES-ACT2 Power Plant with DCLL Blanket", Fusion Science and Technology, 72, pp. 17-40 (July 4, 2017) Abstract BibTeX The ARIES team has examined a multitude of fusion concepts over a period of 25 years. In recent years, the team wrapped up the Advanced Research, Innovation, and Evaluation Study (ARIES) series by completing the detailed design of the ARIES–Advanced and Conservative Tokamak (ARIES-ACT2) power plant—a plant with conservative physics and technology, representing a tokamak with reduced-activation ferritic/martensitic (RAFM) structure and dual-coolant lead-lithium blanket. The integration of nuclear assessments (neutronics, shielding, and activation) is an essential element to ARIES-ACT2 success. This paper highlights the design philosophy of in-vessel components and characterizes several nuclear-related issues that have been addressed during the course of the study to improve the ARIES-ACT2 design: sufficient breeding of tritium to fuel the plasma, well-optimized in-vessel components that satisfy all design requirements and guarantee the shielding functionality of its radial/vertical builds, survivability of low-activation/radiation-resistant structural materials in 14-MeV neutron environment, activation concerns for RAFM and corrosion-resistant oxide-dispersion-strengthened alloys, and an integral approach to handle the mildly radioactive materials during operation and after decommissioning. @article{el-guebaly_design_2017, title = {Design and {Evaluation} of {Nuclear} {System} for {ARIES}-{ACT2} {Power} {Plant} with {DCLL} {Blanket}}, volume = {72}, issn = {1536-1055}, url = {https://doi.org/10.1080/15361055.2016.1273669}, doi = {10.1080/15361055.2016.1273669}, abstract = {The ARIES team has examined a multitude of fusion concepts over a period of 25 years. In recent years, the team wrapped up the Advanced Research, Innovation, and Evaluation Study (ARIES) series by completing the detailed design of the ARIES–Advanced and Conservative Tokamak (ARIES-ACT2) power plant—a plant with conservative physics and technology, representing a tokamak with reduced-activation ferritic/martensitic (RAFM) structure and dual-coolant lead-lithium blanket. The integration of nuclear assessments (neutronics, shielding, and activation) is an essential element to ARIES-ACT2 success. This paper highlights the design philosophy of in-vessel components and characterizes several nuclear-related issues that have been addressed during the course of the study to improve the ARIES-ACT2 design: sufficient breeding of tritium to fuel the plasma, well-optimized in-vessel components that satisfy all design requirements and guarantee the shielding functionality of its radial/vertical builds, survivability of low-activation/radiation-resistant structural materials in 14-MeV neutron environment, activation concerns for RAFM and corrosion-resistant oxide-dispersion-strengthened alloys, and an integral approach to handle the mildly radioactive materials during operation and after decommissioning.}, number = {1}, urldate = {2018-04-05}, journal = {Fusion Science and Technology}, author = {El-Guebaly, L. and Mynsberge, L. and Davis, A. and D’Angelo, C. and Rowcliffe, A. and Pint, B. and Team, ARIES-ACT}, month = jul, year = {2017}, keywords = {Activation analysis, DCLL blanket, neutronics, product}, pages = {17--40}, } Elliott D. Biondo, Paul P.H. Wilson, "Transmutation Approximations for the Application of Hybrid Monte Carlo/Deterministic Neutron Transport to Shutdown Dose Rate Analysis", Nuclear Science and Engineering, 187, pp. 27--48 (July 2017) Abstract BibTeX In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 ± 100 relative to global variance reduction with the Forward Weighted (FW)-CADIS method and 9 ± 5 · 10^4 relative to analog. This work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis. @article{biondo_transmutation_2017, title = {Transmutation {Approximations} for the {Application} of {Hybrid} {Monte} {Carlo}/{Deterministic} {Neutron} {Transport} to {Shutdown} {Dose} {Rate} {Analysis}}, volume = {187}, url = {http://www.tandfonline.com/doi/abs/10.1080/00295639.2016.1275848}, abstract = {In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 ± 100 relative to global variance reduction with the Forward Weighted (FW)-CADIS method and 9 ± 5 · 10{\textasciicircum}4 relative to analog. This work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.}, number = {1}, journal = {Nuclear Science and Engineering}, author = {Biondo, Elliott D. and Wilson, Paul P.H.}, month = jul, year = {2017}, keywords = {CNERG:HK20 Final Report, product}, pages = {27--48}, } Moataz Harb, Paul P.H. Wilson, Andrew Davis, "The Effect of Constructed Mesh-Based Fluxes on Shutdown Dose Rate Calculations in Fusion Energy Systems", Transactions of the American Nuclear Society, 117, pp. 1216-1219 (11/2017) BibTeX @article{harb_effect_2017, title = {The {Effect} of {Constructed} {Mesh}-{Based} {Fluxes} on {Shutdown} {Dose} {Rate} {Calculations} in {Fusion} {Energy} {Systems}}, volume = {117}, number = {1}, journal = {Transactions of the American Nuclear Society}, author = {Harb, Moataz and Wilson, Paul P.H. and Davis, Andrew}, month = nov, year = {2017}, keywords = {product}, pages = {1216--1219}, } Chelsea A D’Angelo, Andrew Davis, Paul P H Wilson, "Recovering Topology of Nested Volumes Represented by Single Closed Surfaces", Transactions of the American Nuclear Society, 116, pp. (2017) BibTeX @article{dangelo_recovering_2017, title = {Recovering {Topology} of {Nested} {Volumes} {Represented} by {Single} {Closed} {Surfaces}}, volume = {116}, language = {en}, journal = {Transactions of the American Nuclear Society}, author = {D’Angelo, Chelsea A and Davis, Andrew and Wilson, Paul P H}, year = {2017}, } C. E. Kessel, J. P. Blanchard, A. Davis, L. El-Guebaly, L. M. Garrison, et al, "Overview of the fusion nuclear science facility, a credible break-in step on the path to fusion energy", Fusion Engineering and Design, , pp. (2017) Abstract BibTeX The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R = 4.8 m, a = 1.2 m, IP = 7.9 MA, BT = 7.5 T, βN < 2.7, n/nGr = 0.9, fBS = 0.52, q95 = 6.0, H98 ∼1.0, and Q = 4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25 years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations. @article{kessel_overview_2017, title = {Overview of the fusion nuclear science facility, a credible break-in step on the path to fusion energy}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S0920379617306257}, doi = {10.1016/j.fusengdes.2017.05.081}, abstract = {The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R = 4.8 m, a = 1.2 m, IP = 7.9 MA, BT = 7.5 T, βN \< 2.7, n/nGr = 0.9, fBS = 0.52, q95 = 6.0, H98 ∼1.0, and Q = 4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25 years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R\&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R\&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R\&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.}, journal = {Fusion Engineering and Design}, author = {Kessel, C. E. and Blanchard, J. P. and Davis, A. and El-Guebaly, L. and Garrison, L. M. and Ghoniem, N. M. and Humrickhouse, P. W. and Huang, Y. and Katoh, Y. and Khodak, A. and Marriott, E. P. and Malang, S. and Morley, N. B. and Neilson, G. H. and Rapp, J. and Rensink, M. E. and Rognlien, T. D. and Rowcliffe, A. F. and Smolentsev, S. and Snead, L. L. and Tillack, M. S. and Titus, P. and Waganer, L. M. and Wallace, G. M. and Wukitch, S. J. and Ying, A. and Young, K. and Zhai, Y.}, year = {2017}, keywords = {Blanket, Divertor, Fusion, Nuclear, Plasma, Systems}, } Matthew J. Gidden, Paul P. H. Wilson, "A methodology for determining the dynamic exchange of resources in nuclear fuel cycle simulation", Nuclear Engineering and Design, 310, pp. 378-394 (December 15, 2016) Abstract BibTeX Simulation of the nuclear fuel cycle can be performed using a wide range of techniques and methodologies. Past efforts have focused on specific fuel cycles or reactor technologies. The CYCLUS fuel cycle simulator seeks to separate the design of the simulation from the fuel cycle or technologies of interest. In order to support this separation, a robust supply–demand communication and solution framework is required. Accordingly an agent-based supply-chain framework, the Dynamic Resource Exchange (DRE), has been designed implemented in CYCLUS. It supports the communication of complex resources, namely isotopic compositions of nuclear fuel, between fuel cycle facilities and their managers (e.g., institutions and regions). Instances of supply and demand are defined as an optimization problem and solved for each timestep. Importantly, the DRE allows each agent in the simulation to independently indicate preference for specific trading options in order to meet both physics requirements and satisfy constraints imposed by potential socio-political models. To display the variety of possible simulations that the DRE enables, example scenarios are formulated and described. Important features include key fuel-cycle facility outages, introduction of external recycled fuel sources (similar to the current mixed oxide (MOX) fuel fabrication facility in the United States), and nontrivial interactions between fuel cycles existing in different regions. @article{gidden_methodology_2016, title = {A methodology for determining the dynamic exchange of resources in nuclear fuel cycle simulation}, volume = {310}, issn = {0029-5493}, url = {http://www.sciencedirect.com/science/article/pii/S0029549316304101}, doi = {10.1016/j.nucengdes.2016.10.029}, abstract = {Simulation of the nuclear fuel cycle can be performed using a wide range of techniques and methodologies. Past efforts have focused on specific fuel cycles or reactor technologies. The CYCLUS fuel cycle simulator seeks to separate the design of the simulation from the fuel cycle or technologies of interest. In order to support this separation, a robust supply–demand communication and solution framework is required. Accordingly an agent-based supply-chain framework, the Dynamic Resource Exchange (DRE), has been designed implemented in CYCLUS. It supports the communication of complex resources, namely isotopic compositions of nuclear fuel, between fuel cycle facilities and their managers (e.g., institutions and regions). Instances of supply and demand are defined as an optimization problem and solved for each timestep. Importantly, the DRE allows each agent in the simulation to independently indicate preference for specific trading options in order to meet both physics requirements and satisfy constraints imposed by potential socio-political models. To display the variety of possible simulations that the DRE enables, example scenarios are formulated and described. Important features include key fuel-cycle facility outages, introduction of external recycled fuel sources (similar to the current mixed oxide (MOX) fuel fabrication facility in the United States), and nontrivial interactions between fuel cycles existing in different regions.}, urldate = {2017-01-11}, journal = {Nuclear Engineering and Design}, author = {Gidden, Matthew J. and Wilson, Paul P. H.}, month = dec, year = {2016}, keywords = {Agent-based modeling, NEWTON, Nuclear fuel cycle, Optimization}, pages = {378--394}, } Robert W. Carlsen, Paul P. H. Wilson, "Challenging Fuel Cycle Modeling Assumptions: Facility and Time-Step Discretization Effects", Nuclear Technology, 195, pp. (2016-09-01) BibTeX @article{carlsen_challenging_2016, title = {Challenging {Fuel} {Cycle} {Modeling} {Assumptions}: {Facility} and {Time}-{Step} {Discretization} {Effects}}, volume = {195}, issn = {00295450}, shorttitle = {Challenging {Fuel} {Cycle} {Modeling} {Assumptions}}, url = {http://www.ans.org/pubs/journals/nt/a_38930}, doi = {10.13182/NT15-138}, number = {3}, urldate = {2016-08-11}, journal = {Nuclear Technology}, author = {Carlsen, Robert W. and Wilson, Paul P. H.}, month = sep, year = {2016}, } Kathryn D. Huff, Matthew J. Gidden, Robert W. Carlsen, Robert R. Flanagan, Meghan B. McGarry, et al, "Fundamental concepts in the Cyclus nuclear fuel cycle simulation framework", Advances in Engineering Software, 94, pp. 46-59 (April 2016) Abstract BibTeX As nuclear power expands, technical, economic, political, and environmental analyses of nuclear fuel cycles by simulators increase in importance. To date, however, current tools are often fleet-based rather than discrete and restrictively licensed rather than open source. Each of these choices presents a challenge to modeling fidelity, generality, efficiency, robustness, and scientific transparency. The Cyclus nuclear fuel cycle simulator framework and its modeling ecosystem incorporate modern insights from simulation science and software architecture to solve these problems so that challenges in nuclear fuel cycle analysis can be better addressed. A summary of the Cyclus fuel cycle simulator framework and its modeling ecosystem are presented. Additionally, the implementation of each is discussed in the context of motivating challenges in nuclear fuel cycle simulation. Finally, the current capabilities of Cyclus are demonstrated for both open and closed fuel cycles. @article{huff_fundamental_2016, title = {Fundamental concepts in the {Cyclus} nuclear fuel cycle simulation framework}, volume = {94}, issn = {0965-9978}, url = {http://www.sciencedirect.com/science/article/pii/S0965997816300229}, doi = {10.1016/j.advengsoft.2016.01.014}, abstract = {As nuclear power expands, technical, economic, political, and environmental analyses of nuclear fuel cycles by simulators increase in importance. To date, however, current tools are often fleet-based rather than discrete and restrictively licensed rather than open source. Each of these choices presents a challenge to modeling fidelity, generality, efficiency, robustness, and scientific transparency. The Cyclus nuclear fuel cycle simulator framework and its modeling ecosystem incorporate modern insights from simulation science and software architecture to solve these problems so that challenges in nuclear fuel cycle analysis can be better addressed. A summary of the Cyclus fuel cycle simulator framework and its modeling ecosystem are presented. Additionally, the implementation of each is discussed in the context of motivating challenges in nuclear fuel cycle simulation. Finally, the current capabilities of Cyclus are demonstrated for both open and closed fuel cycles.}, urldate = {2016-07-31}, journal = {Advances in Engineering Software}, author = {Huff, Kathryn D. and Gidden, Matthew J. and Carlsen, Robert W. and Flanagan, Robert R. and McGarry, Meghan B. and Opotowsky, Arrielle C. and Schneider, Erich A. and Scopatz, Anthony M. and Wilson, Paul P. H.}, month = apr, year = {2016}, keywords = {Agent based modeling, NEWTON, Nuclear engineering, Nuclear fuel cycle, Object orientation, Simulation, Systems analysis}, pages = {46--59}, } Elliott D. Biondo, Andrew Davis, Paul P.H. Wilson, "Shutdown Dose Rate Analysis with CAD Geometry, Cartesian/Tetrahedral Mesh, and Advanced Variance Reduction", Fusion Engineering and Design, 106, pp. 77-84 (05/2016) Abstract BibTeX In fusion energy systems (FES) high-energy neutrons born from burning plasma activate system components to form radionuclides. The biological dose rate that results from photons emitted by these radionuclides after shutdown—the shutdown dose rate (SDR)—must be quantified for maintenance planning. This can be done using the Rigorous Two-Step (R2S) method, which involves separate neutron and photon transport calculations, coupled by a nuclear inventory analysis code. The geometric complexity and highly attenuating configuration of FES motivates the use of CAD geometry and advanced variance reduction for this analysis. An R2S workflow has been created with the new capability of performing SDR analysis directly from CAD geometry with Cartesian or tetrahedral meshes and with biased photon source sampling, enabling the use of the Consistent Adjoint Driven Importance Sampling (CADIS) variance reduction technique. This workflow has been validated with the Frascati Neutron Generator (FNG)-ITER SDR benchmark using both Cartesian and tetrahedral meshes and both unbiased and biased photon source sampling. All results are within 20.4% of experimental values, which constitutes satisfactory agreement. Photon transport using CADIS is demonstrated to yield speedups as high as 8.5·10^5 for problems using the FNG geometry. @article{biondo_shutdown_2016, title = {Shutdown {Dose} {Rate} {Analysis} with {CAD} {Geometry}, {Cartesian}/{Tetrahedral} {Mesh}, and {Advanced} {Variance} {Reduction}}, volume = {106}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S0920379616302009}, doi = {http://dx.doi.org/10.1016/j.fusengdes.2016.03.004}, abstract = {In fusion energy systems (FES) high-energy neutrons born from burning plasma activate system components to form radionuclides. The biological dose rate that results from photons emitted by these radionuclides after shutdown—the shutdown dose rate (SDR)—must be quantified for maintenance planning. This can be done using the Rigorous Two-Step (R2S) method, which involves separate neutron and photon transport calculations, coupled by a nuclear inventory analysis code. The geometric complexity and highly attenuating configuration of FES motivates the use of CAD geometry and advanced variance reduction for this analysis. An R2S workflow has been created with the new capability of performing SDR analysis directly from CAD geometry with Cartesian or tetrahedral meshes and with biased photon source sampling, enabling the use of the Consistent Adjoint Driven Importance Sampling (CADIS) variance reduction technique. This workflow has been validated with the Frascati Neutron Generator (FNG)-ITER SDR benchmark using both Cartesian and tetrahedral meshes and both unbiased and biased photon source sampling. All results are within 20.4\% of experimental values, which constitutes satisfactory agreement. Photon transport using CADIS is demonstrated to yield speedups as high as 8.5·10{\textasciicircum}5 for problems using the FNG geometry.}, journal = {Fusion Engineering and Design}, author = {Biondo, Elliott D. and Davis, Andrew and Wilson, Paul P.H.}, month = may, year = {2016}, keywords = {CNERG:HK20 Final Report, product}, pages = {77--84}, } Stuart R. Slattery, Thomas M. Evans, Paul P. H. Wilson, "A spectral analysis of the domain decomposed Monte Carlo method for linear systems", Nuclear Engineering and Design, 295, pp. 632-638 (12/15/2015) Abstract BibTeX The domain decomposed behavior of the adjoint Neumann-Ulam Monte Carlo method for solving linear systems is analyzed using the spectral properties of the linear operator. Relationships for the average length of the adjoint random walks, a measure of convergence speed and serial performance, are made with respect to the eigenvalues of the linear operator. In addition, relationships for the effective optical thickness of a domain in the decomposition are presented based on the spectral analysis and diffusion theory. Using the effective optical thickness, the Wigner rational approximation and the mean chord approximation are applied to estimate the leakage fraction of random walks from a domain in the decomposition as a measure of parallel performance and potential communication costs. The one-speed, two-dimensional neutron diffusion equation is used as a model problem in numerical experiments to test the models for symmetric operators with spectral qualities similar to light water reactor problems. In general, the derived approximations show good agreement with random walk lengths and leakage fractions computed by the numerical experiments. @article{slattery_spectral_2015, title = {A spectral analysis of the domain decomposed {Monte} {Carlo} method for linear systems}, volume = {295}, issn = {0029-5493}, url = {http://www.sciencedirect.com/science/article/pii/S0029549315003271}, doi = {10.1016/j.nucengdes.2015.07.054}, abstract = {The domain decomposed behavior of the adjoint Neumann-Ulam Monte Carlo method for solving linear systems is analyzed using the spectral properties of the linear operator. Relationships for the average length of the adjoint random walks, a measure of convergence speed and serial performance, are made with respect to the eigenvalues of the linear operator. In addition, relationships for the effective optical thickness of a domain in the decomposition are presented based on the spectral analysis and diffusion theory. Using the effective optical thickness, the Wigner rational approximation and the mean chord approximation are applied to estimate the leakage fraction of random walks from a domain in the decomposition as a measure of parallel performance and potential communication costs. The one-speed, two-dimensional neutron diffusion equation is used as a model problem in numerical experiments to test the models for symmetric operators with spectral qualities similar to light water reactor problems. In general, the derived approximations show good agreement with random walk lengths and leakage fractions computed by the numerical experiments.}, urldate = {2015-09-09}, journal = {Nuclear Engineering and Design}, author = {Slattery, Stuart R. and Evans, Thomas M. and Wilson, Paul P. H.}, month = dec, year = {2015}, pages = {632--638}, } Robert W. Carlsen, Wilson, Paul P.H., "Fast Finding of Fast Transitions to Fast Reactor Fuel Cycles with Cyclus", Transactions of the American Nuclear Society, 113, pp. 364-367 (November 2015) Abstract BibTeX While many potential fuel cycles have been evaluated for their equilibrium properties, there has been less investigation into the properties of transitions from the current once-through light water reactor (LWR) fuel cycle to other candidate fuel cycles. This work describes a transition analysis performed using black-box optimization techniques coupled with the Cyclus fuel cycle simulator on large scale computing resources. A transition deployment schedule for thermal and fast reactors was computed by wrapping Cyclus simulations inside an external optimizer. The resulting deployments provide an approximate optimal solution for the fastest way to transition to an EG23-like fuel cycle without forced early retirement of reactors (for the utilized objective function). Although only reactor deployments are looked at here, the optimization workflow developed is flexible enough to handle deployment of an arbitrary number of facility types. The optimization runs were performed on the HT Condor high throughput computing infrastructure at University of Wisconsin-Madison. Optimization tools were developed that are robust against failures and instability in the cluster environment. The optimizer used is a hybrid of particle swarm and modified pattern search algorithms based on PSwarm. It is robust against failed simulations and objective function evaluations and is easy to parallelize. @article{carlsen_fast_2015, title = {Fast {Finding} of {Fast} {Transitions} to {Fast} {Reactor} {Fuel} {Cycles} with {Cyclus}}, volume = {113}, abstract = {While many potential fuel cycles have been evaluated for their equilibrium properties, there has been less investigation into the properties of transitions from the current once-through light water reactor (LWR) fuel cycle to other candidate fuel cycles. This work describes a transition analysis performed using black-box optimization techniques coupled with the Cyclus fuel cycle simulator on large scale computing resources. A transition deployment schedule for thermal and fast reactors was computed by wrapping Cyclus simulations inside an external optimizer. The resulting deployments provide an approximate optimal solution for the fastest way to transition to an EG23-like fuel cycle without forced early retirement of reactors (for the utilized objective function). Although only reactor deployments are looked at here, the optimization workflow developed is flexible enough to handle deployment of an arbitrary number of facility types. The optimization runs were performed on the HT Condor high throughput computing infrastructure at University of Wisconsin-Madison. Optimization tools were developed that are robust against failures and instability in the cluster environment. The optimizer used is a hybrid of particle swarm and modified pattern search algorithms based on PSwarm. It is robust against failed simulations and objective function evaluations and is easy to parallelize.}, journal = {Transactions of the American Nuclear Society}, author = {Carlsen, Robert W. and Wilson, Paul P.H.}, month = nov, year = {2015}, keywords = {NEWTON}, pages = {364--367}, } Bastian Weinhorst, Ulrich Fischer, Lei Lu, Yuefeng Qiu, Paul Wilson, "Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations", Fusion Engineering and Design, 98-99, pp. 2094-2097 (October 1, 2015) Abstract BibTeX Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user. @article{weinhorst_comparative_2015, series = {Proceedings of the 28th {Symposium} {On} {Fusion} {Technology} ({SOFT}-28)}, title = {Comparative assessment of different approaches for the use of {CAD} geometry in {Monte} {Carlo} transport calculations}, volume = {98-99}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S0920379615300739}, doi = {10.1016/j.fusengdes.2015.06.042}, abstract = {Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.}, urldate = {2018-04-09}, journal = {Fusion Engineering and Design}, author = {Weinhorst, Bastian and Fischer, Ulrich and Lu, Lei and Qiu, Yuefeng and Wilson, Paul}, month = oct, year = {2015}, keywords = {CAD, DAGMC, MCNP, McCad, Neutronics}, pages = {2094--2097}, } Tim Bohm, Andrew Davis, Mohamed Sawan, Edward Marriott, Paul Wilson, et al, "Detailed 3-D nuclear analysis of ITER outboard blanket modules", Fusion Engineering and Design, 96-97, pp. 222-226 (2015-10-01) Abstract BibTeX In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded. @article{bohm_detailed_2015, series = {Proceedings of the 28th {Symposium} {On} {Fusion} {Technology} ({SOFT}-28)}, title = {Detailed 3-{D} nuclear analysis of {ITER} outboard blanket modules}, volume = {96-97}, issn = {0920-3796}, url = {https://www.sciencedirect.com/science/article/pii/S0920379615300351}, doi = {10.1016/j.fusengdes.2015.06.004}, abstract = {In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80\%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45\%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.}, urldate = {2023-10-29}, journal = {Fusion Engineering and Design}, author = {Bohm, Tim and Davis, Andrew and Sawan, Mohamed and Marriott, Edward and Wilson, Paul and Ulrickson, Michael and Bullock, James}, month = oct, year = {2015}, keywords = {Blanket modules, CAD, ITER, MCNP}, pages = {222--226}, } Ahmad M. Ibrahim, Paul P. H. Wilson, Mohamed E. Sawan, Scott W. Mosher, Douglas E. Peplow, et al, "Automatic Mesh Adaptivity for Hybrid Monte Carlo/Deterministic Neutronics Modeling of Difficult Shielding Problems", Nuclear Science and Engineering, 181, pp. 48-59 (2015/09/01) BibTeX @article{ibrahim_automatic_2015, title = {Automatic {Mesh} {Adaptivity} for {Hybrid} {Monte} {Carlo}/{Deterministic} {Neutronics} {Modeling} of {Difficult} {Shielding} {Problems}}, volume = {181}, url = {http://epubs.ans.org/?a=37494}, doi = {dx.doi.org/10.13182/NSE14-94}, number = {1}, urldate = {2016-10-26}, journal = {Nuclear Science and Engineering}, author = {Ibrahim, Ahmad M. and Wilson, Paul P. H. and Sawan, Mohamed E. and Mosher, Scott W. and Peplow, Douglas E. and Wagner, John C. and Evans, Thomas M. and Grove, Robert E.}, month = sep, year = {2015}, pages = {48--59}, } Anthony M. Scopatz, Cameron R. Bates, Paul P.H. Wilson, "Binary Formulation of Decay Equations", Transactions of the American Nuclear Society, 112, pp. 71-74 (June 2015) BibTeX @article{scopatz_binary_2015, title = {Binary {Formulation} of {Decay} {Equations}}, volume = {112}, journal = {Transactions of the American Nuclear Society}, author = {Scopatz, Anthony M. and Bates, Cameron R. and Wilson, Paul P.H.}, month = jun, year = {2015}, pages = {71--74}, } C. E. Kessel, M. S. Tillack, F. Najmabadi, F. M. Poli, K. Ghantous, et al, "The ARIES Advanced and Conservative Tokamak Power Plant Study", Fusion Science and Technology, 67, pp. 1-21 (January 1, 2015) Abstract BibTeX Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5,a βtotal N of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m2. The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced-activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βtotal N of 2.5, an H98 of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here. @article{kessel_aries_2015, title = {The {ARIES} {Advanced} and {Conservative} {Tokamak} {Power} {Plant} {Study}}, volume = {67}, issn = {1536-1055}, url = {https://doi.org/10.13182/FST14-794}, doi = {10.13182/FST14-794}, abstract = {Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58\% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5,a βtotal N of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m2. The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced-activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45\%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βtotal N of 2.5, an H98 of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.}, number = {1}, urldate = {2018-04-05}, journal = {Fusion Science and Technology}, author = {Kessel, C. E. and Tillack, M. S. and Najmabadi, F. and Poli, F. M. and Ghantous, K. and Gorelenkov, N. and Wang, X. R. and Navaei, D. and Toudeshki, H. H. and Koehly, C. and EL-Guebaly, L. and Blanchard, J. P. and Martin, C. J. and Mynsburge, L. and Humrickhouse, P. and Rensink, M. E. and Rognlien, T. D. and Yoda, M. and Abdel-Khalik, S. I. and Hageman, M. D. and MILLS, B. H. and Rader, J. D. and Sadowski, D. L. and Snyder, P. B. and John, H. ST and Turnbull, A. D. and Waganer, L. M. and MALANG, S. and Rowcliffe, A. F.}, month = jan, year = {2015}, pages = {1--21}, } Anthony Scopatz, Arrielle Opotowsky, Paul P. H. Wilson, "Cymetric - A Fuel Cycle Metrics Tool for Cyclus", Transactions of the American Nuclear Society, 112, pp. 81-84 (6/2015) BibTeX @article{scopatz_cymetric_2015, title = {Cymetric - {A} {Fuel} {Cycle} {Metrics} {Tool} for {Cyclus}}, volume = {112}, number = {1}, journal = {Transactions of the American Nuclear Society}, author = {Scopatz, Anthony and Opotowsky, Arrielle and Wilson, Paul P. H.}, month = jun, year = {2015}, pages = {81--84}, } Elliott Biondo, Andrew Davis, Anthony Scopatz, Paul P.H. Wilson, "Rigorous Two-Step Activation for Fusion Systems with PyNE", Transactions of the American Nuclear Society, 112, pp. 617-620 (2015) BibTeX @article{biondo_rigorous_2015, series = {Best of {Radiation} {Protection} and {Shielding} {Division} 2014---{II}}, title = {Rigorous {Two}-{Step} {Activation} for {Fusion} {Systems} with {PyNE}}, volume = {112}, journal = {Transactions of the American Nuclear Society}, author = {Biondo, Elliott and Davis, Andrew and Scopatz, Anthony and Wilson, Paul P.H.}, year = {2015}, keywords = {CNERG:HK20 Final Report, product}, pages = {617--620}, } Elliott Biondo, Anthony Scopatz, Matthew Gidden, Rachel Slaybaugh, Cameron Bates, et al, "Quality Assurance within the PyNE Open Source Toolkit", Transactions of the American Nuclear Society, 111, pp. (November 9-13, 2014) BibTeX @article{biondo_quality_2014, title = {Quality {Assurance} within the {PyNE} {Open} {Source} {Toolkit}}, volume = {111}, url = {https://github.com/pyne/ans-winter-2014-vnv}, journal = {Transactions of the American Nuclear Society}, author = {Biondo, Elliott and Scopatz, Anthony and Gidden, Matthew and Slaybaugh, Rachel and Bates, Cameron and WIlson, Paul P.H.}, month = nov, year = {2014}, } Robert W. Carlsen, Matthew J. Gidden, Paul P.H. Wilson, "Deployment Optimization with the CYCLUS Fuel Cycle Simulator", Transactions of the American Nuclear Society, 111, pp. 241-244 (November 2014) BibTeX @article{carlsen_deployment_2014, title = {Deployment {Optimization} with the {CYCLUS} {Fuel} {Cycle} {Simulator}}, volume = {111}, journal = {Transactions of the American Nuclear Society}, author = {Carlsen, Robert W. and Gidden, Matthew J. and Wilson, Paul P.H.}, month = nov, year = {2014}, note = {DOI link for code, methods, etc: http://dx.doi.org/10.6084/m9.figshare.1086284}, keywords = {NEWTON}, pages = {241--244}, } T. D. Bohm, M. E. Sawan, E. P. Marriott, P. P. H. Wilson, M. Ulrickson, et al, "Detailed 3-D nuclear analysis of ITER blanket modules", Fusion Engineering and Design, 89, pp. 1954-1958 (October 2014) Abstract BibTeX In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded. @article{bohm_detailed_2014, series = {Proceedings of the 11th {International} {Symposium} on {Fusion} {Nuclear} {Technology}-11 ({ISFNT}-11) {Barcelona}, {Spain}, 15-20 {September}, 2013}, title = {Detailed 3-{D} nuclear analysis of {ITER} blanket modules}, volume = {89}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S092037961400057X}, doi = {10.1016/j.fusengdes.2014.01.056}, abstract = {In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20\%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.}, number = {9–10}, urldate = {2014-08-08}, journal = {Fusion Engineering and Design}, author = {Bohm, T. D. and Sawan, M. E. and Marriott, E. P. and Wilson, P. P. H. and Ulrickson, M. and Bullock, J.}, month = oct, year = {2014}, keywords = {Blanket module, CAD, ITER, MCNP}, pages = {1954--1958}, } Ahmad M. Ibrahim, Paul P. Wilson, Mohamed E. Sawan, Scott W. Mosher, Douglas E. Peplow, et al, "Assessment of fusion facility dose rate map using mesh adaptivity enhancements of hybrid Monte Carlo/deterministic techniques", Fusion Engineering and Design, 89, pp. 1875-1879 (October 2014) Abstract BibTeX Three mesh adaptivity algorithms were developed to facilitate and expedite the use of the CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques in accurate full-scale neutronics simulations of fusion energy systems with immense sizes and complicated geometries. First, a macromaterial approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm decouples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility and resulted in a 23.3% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation. Additionally, because of the significant increase in the efficiency of FW-CADIS simulations, the three algorithms enabled this difficult calculation to be accurately solved on a regular computer cluster, eliminating the need for a world-class super computer. @article{ibrahim_assessment_2014, series = {Proceedings of the 11th {International} {Symposium} on {Fusion} {Nuclear} {Technology}-11 ({ISFNT}-11) {Barcelona}, {Spain}, 15-20 {September}, 2013}, title = {Assessment of fusion facility dose rate map using mesh adaptivity enhancements of hybrid {Monte} {Carlo}/deterministic techniques}, volume = {89}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S0920379614001434}, doi = {10.1016/j.fusengdes.2014.02.046}, abstract = {Three mesh adaptivity algorithms were developed to facilitate and expedite the use of the CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques in accurate full-scale neutronics simulations of fusion energy systems with immense sizes and complicated geometries. First, a macromaterial approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm decouples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility and resulted in a 23.3\% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation. Additionally, because of the significant increase in the efficiency of FW-CADIS simulations, the three algorithms enabled this difficult calculation to be accurately solved on a regular computer cluster, eliminating the need for a world-class super computer.}, number = {9–10}, urldate = {2014-08-08}, journal = {Fusion Engineering and Design}, author = {Ibrahim, Ahmad M. and Wilson, Paul P. and Sawan, Mohamed E. and Mosher, Scott W. and Peplow, Douglas E. and Grove, Robert E.}, month = oct, year = {2014}, keywords = {Hybrid Monte Carlo/deterministic, ITER prompt dose, Neutronics shielding}, pages = {1875--1879}, } Po Hu, Paul P. H. Wilson, "Core Flow Distribution from Coupled Supercritical Water Reactor Analysis", Science and Technology of Nuclear Installations, 2014, pp. (2014) Abstract BibTeX This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal. @article{hu_core_2014, title = {Core {Flow} {Distribution} from {Coupled} {Supercritical} {Water} {Reactor} {Analysis}}, volume = {2014}, url = {https://www.hindawi.com/journals/stni/2014/178129/}, doi = {http://dx.doi.org/10.1155/2014/178129}, abstract = {This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.}, language = {en}, urldate = {2018-02-17}, journal = {Science and Technology of Nuclear Installations}, author = {Hu, Po and Wilson, Paul P. H.}, year = {2014}, doi = {10.1155/2014/178129}, } Po Hu, Paul Wilson, "Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor", Science and Technology of Nuclear Installations, 2014, pp. (2014) Abstract BibTeX The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized. @article{hu_code_2014, title = {Code {Development} in {Coupled} {PARCS}/{RELAP5} for {Supercritical} {Water} {Reactor}}, volume = {2014}, url = {https://www.hindawi.com/journals/stni/2014/286434/}, doi = {http://dx.doi.org/10.1155/2014/286434}, abstract = {The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.}, language = {en}, urldate = {2018-02-17}, journal = {Science and Technology of Nuclear Installations}, author = {Hu, Po and Wilson, Paul}, year = {2014}, doi = {10.1155/2014/286434}, } K. L. Dunn, P. P. H. Wilson, "MNCP5 vs KDE: Direct Method for Mesh Tally Comparisons", Transactions of the American Nuclear Society, 109, pp. 691-694 (Nov 2013) BibTeX @article{dunn_mncp5_2013, title = {{MNCP5} vs {KDE}: {Direct} {Method} for {Mesh} {Tally} {Comparisons}}, volume = {109}, number = {1}, journal = {Transactions of the American Nuclear Society}, author = {Dunn, K. L. and Wilson, P. P. H.}, month = nov, year = {2013}, pages = {691--694}, } Min Cheol Han, Chan Hyeong Kim, Jong Hwi Jeong, Yeon Soo Yeom, SungHoon Kim, et al, "DagSolid: a new Geant4 solid class for fast simulation in polygon-mesh geometry", Physics in Medicine and Biology, 58, pp. 4595 (2013-07-07) Abstract BibTeX Even though a computer-aided design (CAD)-based geometry can be directly implemented in Geant4 as polygon-mesh using the G4TessellatedSolid class, the computation speed becomes very slow, especially when the geometry is composed of a large number of facets. To address this problem, in the present study, a new Geant4 solid class, named DagSolid, was developed based on the direct accelerated geometry for the Monte Carlo (DAGMC) library which provides the ray-tracing acceleration algorithm functions. To develop the DagSolid class, the new solid class was derived from the G4VSolid class, and its ray-tracing functions were linked to the corresponding functions of the DAGMC library. The results of this study show that the use of the DagSolid class drastically improves the computation speed. The improvement was more significant when there were more facets, meaning that the DagSolid class can be used more effectively for complicated geometries with many facets than for simple geometries. The maximum difference of computation speed was 1562 and 680 times for Geantino and ChargedGeantino, respectively. For real particles (gammas, electrons, neutrons, and protons), the difference of computation speed was less significant, but still was within the range of 53–685 times depending on the type of beam particles simulated. @article{han_dagsolid:_2013, title = {{DagSolid}: a new {Geant4} solid class for fast simulation in polygon-mesh geometry}, volume = {58}, issn = {0031-9155}, shorttitle = {{DagSolid}}, url = {http://iopscience.iop.org/0031-9155/58/13/4595}, doi = {10.1088/0031-9155/58/13/4595}, abstract = {Even though a computer-aided design (CAD)-based geometry can be directly implemented in Geant4 as polygon-mesh using the G4TessellatedSolid class, the computation speed becomes very slow, especially when the geometry is composed of a large number of facets. To address this problem, in the present study, a new Geant4 solid class, named DagSolid, was developed based on the direct accelerated geometry for the Monte Carlo (DAGMC) library which provides the ray-tracing acceleration algorithm functions. To develop the DagSolid class, the new solid class was derived from the G4VSolid class, and its ray-tracing functions were linked to the corresponding functions of the DAGMC library. The results of this study show that the use of the DagSolid class drastically improves the computation speed. The improvement was more significant when there were more facets, meaning that the DagSolid class can be used more effectively for complicated geometries with many facets than for simple geometries. The maximum difference of computation speed was 1562 and 680 times for Geantino and ChargedGeantino, respectively. For real particles (gammas, electrons, neutrons, and protons), the difference of computation speed was less significant, but still was within the range of 53–685 times depending on the type of beam particles simulated.}, language = {en}, number = {13}, urldate = {2013-06-21}, journal = {Physics in Medicine and Biology}, author = {Han, Min Cheol and Kim, Chan Hyeong and Jeong, Jong Hwi and Yeom, Yeon Soo and Kim, SungHoon and Wilson, Paul P. H. and Apostolakis, John}, month = jul, year = {2013}, pages = {4595}, } Matthew Gidden, Paul Wilson, Anthony Scopatz, "Developing Standardized, Open Benchmarks and a Corresponding Specification Language for the Simulation of Dynamic Fuel Cycles", Transactions of the American Nuclear Society, 108, pp. 127-130 (June 18, 2013) BibTeX @article{gidden_developing_2013, title = {Developing {Standardized}, {Open} {Benchmarks} and a {Corresponding} {Specification} {Language} for the {Simulation} of {Dynamic} {Fuel} {Cycles}}, volume = {108}, number = {1}, journal = {Transactions of the American Nuclear Society}, author = {Gidden, Matthew and Wilson, Paul and Scopatz, Anthony}, month = jun, year = {2013}, pages = {127--130}, } R. N. Slaybaugh, T. M. Evans, G. G. Davidson, P. P. H. Wilson, "Multigrid in energy preconditioner for Krylov solvers", Journal of Computational Physics, 242, pp. 405-419 (June 1, 2013) Abstract BibTeX We have added a new multigrid in energy (MGE) preconditioner to the Denovo discrete-ordinates radiation transport code. This preconditioner takes advantage of a new multilevel parallel decomposition. A multigroup Krylov subspace iterative solver that is decomposed in energy as well as space-angle forms the backbone of the transport solves in Denovo. The space-angle-energy decomposition facilitates scaling to hundreds of thousands of cores. The multigrid in energy preconditioner scales well in the energy dimension and significantly reduces the number of Krylov iterations required for convergence. This preconditioner is well-suited for use with advanced eigenvalue solvers such as Rayleigh Quotient Iteration and Arnoldi. @article{slaybaugh_multigrid_2013, title = {Multigrid in energy preconditioner for {Krylov} solvers}, volume = {242}, issn = {0021-9991}, url = {http://www.sciencedirect.com/science/article/pii/S0021999113001228}, doi = {10.1016/j.jcp.2013.02.012}, abstract = {We have added a new multigrid in energy (MGE) preconditioner to the Denovo discrete-ordinates radiation transport code. This preconditioner takes advantage of a new multilevel parallel decomposition. A multigroup Krylov subspace iterative solver that is decomposed in energy as well as space-angle forms the backbone of the transport solves in Denovo. The space-angle-energy decomposition facilitates scaling to hundreds of thousands of cores. The multigrid in energy preconditioner scales well in the energy dimension and significantly reduces the number of Krylov iterations required for convergence. This preconditioner is well-suited for use with advanced eigenvalue solvers such as Rayleigh Quotient Iteration and Arnoldi.}, urldate = {2017-01-11}, journal = {Journal of Computational Physics}, author = {Slaybaugh, R. N. and Evans, T. M. and Davidson, G. G. and Wilson, P. P. H.}, month = jun, year = {2013}, keywords = {Krylov, Multigrid, Neutron transport, Preconditioning}, pages = {405--419}, } Paul P.H. WIlson, Andrew Davis, Julie Zachman, Kerry L. Dunn, "FluDAG and Other Implementations of the DAGMC Toolkit", Transactions of the American Nuclear Society, 109, pp. 713-716 (11/2013) BibTeX @article{wilson_fludag_2013, title = {{FluDAG} and {Other} {Implementations} of the {DAGMC} {Toolkit}}, volume = {109}, url = {https://github.com/CNERG/publications/tree/ans13w-dagmc}, journal = {Transactions of the American Nuclear Society}, author = {WIlson, Paul P.H. and Davis, Andrew and Zachman, Julie and Dunn, Kerry L.}, month = nov, year = {2013}, pages = {713--716}, } R. Pampin, A. Davis, J. Izquierdo, D. Leichtle, M.J. Loughlin, et al, "Developments and needs in nuclear analysis of fusion technology", Fusion Engineering and Design, , pp. (10/2013) Abstract BibTeX Abstract Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case of fusion technology in current experiments, ITER, next-step devices and power plant studies. Calculations are intricate and computer-intensive, typically requiring detailed geometry models, sophisticated acceleration algorithms, high-performance parallel computations, and coupling of large and complex transport and activation codes and databases. This paper reports progress on some key areas in the development of tools and methods to meet the specific needs of fusion nuclear analyses. In particular, advances in the production and modernisation of reference models, in the preparation and quality assurance of acceleration algorithms and coupling schemes, and in the evaluation and adaptation of alternative transport codes are presented. Emphasis is given to ITER-relevant activities, which are the main driver of advances in the field. Discussion is made of the importance of efforts in these and other areas, considering some of the more pressing needs and requirements. In some cases, they call for a more efficient and coordinated use of the scarce resources available. @article{pampin_developments_2013, title = {Developments and needs in nuclear analysis of fusion technology}, issn = {0920-3796}, url = {http://www.sciencedirect.com/science/article/pii/S0920379613003438}, doi = {10.1016/j.fusengdes.2013.03.049}, abstract = {Abstract Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case of fusion technology in current experiments, ITER, next-step devices and power plant studies. Calculations are intricate and computer-intensive, typically requiring detailed geometry models, sophisticated acceleration algorithms, high-performance parallel computations, and coupling of large and complex transport and activation codes and databases. This paper reports progress on some key areas in the development of tools and methods to meet the specific needs of fusion nuclear analyses. In particular, advances in the production and modernisation of reference models, in the preparation and quality assurance of acceleration algorithms and coupling schemes, and in the evaluation and adaptation of alternative transport codes are presented. Emphasis is given to ITER-relevant activities, which are the main driver of advances in the field. Discussion is made of the importance of efforts in these and other areas, considering some of the more pressing needs and requirements. In some cases, they call for a more efficient and coordinated use of the scarce resources available.}, urldate = {2013-06-10}, journal = {Fusion Engineering and Design}, author = {Pampin, R. and Davis, A. and Izquierdo, J. and Leichtle, D. and Loughlin, M.J. and Sanz, J. and Turner, A. and Villari, R. and Wilson, P.P.H.}, month = oct, year = {2013}, keywords = {Activation, ITER, Neutronics, Nuclear analysis, Shielding}, } Matthew Gidden, Paul Wilson, K. Huff, R. Carlsen, "Once-Through Benchmarks with C YCLUS, a Modular, Open-Source Fuel Cycle Simulator", Transactions of the American Nuclear Society, 107, pp. 264-266 (November 14, 2012) BibTeX @article{gidden_once-through_2012, title = {Once-{Through} {Benchmarks} with {C} {YCLUS}, a {Modular}, {Open}-{Source} {Fuel} {Cycle} {Simulator}}, volume = {107}, number = {1}, journal = {Transactions of the American Nuclear Society}, author = {Gidden, Matthew and Wilson, Paul and Huff, K. and Carlsen, R.}, month = nov, year = {2012}, pages = {264--266}, } Anthony M Scoptaz, Paul K. Romano, Paul P.H. Wilson, Kathryn D. Huff, "PyNE: Python for Nuclear Engineering", Transactions of the American Nuclear Society, 107, pp. 985-987 (November 2012) BibTeX @article{scoptaz_pyne:_2012, title = {{PyNE}: {Python} for {Nuclear} {Engineering}}, volume = {107}, number = {1}, journal = {Transactions of the American Nuclear Society}, author = {Scoptaz, Anthony M and Romano, Paul K. and Wilson, Paul P.H. and Huff, Kathryn D.}, month = nov, year = {2012}, pages = {985--987}, } P. Denholm, J.C. King, C.F. Kutcher, P.P.H. WIlson, "Decarbonizing the electric sector: Combining renewable and nuclear energy using thermal storage", Energy Policy, 44, pp. 301-311 (May 2012) Abstract BibTeX Both renewable and nuclearenergy can provide significant contributions to decarbonizing the electric sector. However, a grid employing large amounts of wind and solar energy requires the balance of the system to be highly flexible to respond to the increased variability of the net load. This makes deployment of conventional nuclear power challenging both due to the technical challenges of plant cycling and economic limits of reduced capacity factor. In the United States nuclear power plants generally provide constant, base load power and are most economic when operated at constant power levels. Operating nuclear power plants in load-following modes decreases the plants' annual energy output and increases the levelized cost of energy, decreasing economic competitiveness. One possible solution is to couple thermalenergystorage to nuclear power plants. This would enable the reactor to remain at nearly constant output, while cycling the electrical generator in response to the variability of the net load. This paper conceptually explores combinations of wind, solar, and nuclear that can provide a large fraction of a system's electricity, assuming the use of thermalenergystorage that would allow nuclear power to provide load following and cycling duty while operating at a constant reactor power output. @article{denholm_decarbonizing_2012, title = {Decarbonizing the electric sector: {Combining} renewable and nuclear energy using thermal storage}, volume = {44}, issn = {0301-4215}, shorttitle = {Decarbonizing the electric sector}, url = {http://www.sciencedirect.com/science/article/pii/S030142151200081X}, doi = {10.1016/j.enpol.2012.01.055}, abstract = {Both renewable and nuclearenergy can provide significant contributions to decarbonizing the electric sector. However, a grid employing large amounts of wind and solar energy requires the balance of the system to be highly flexible to respond to the increased variability of the net load. This makes deployment of conventional nuclear power challenging both due to the technical challenges of plant cycling and economic limits of reduced capacity factor. In the United States nuclear power plants generally provide constant, base load power and are most economic when operated at constant power levels. Operating nuclear power plants in load-following modes decreases the plants' annual energy output and increases the levelized cost of energy, decreasing economic competitiveness. One possible solution is to couple thermalenergystorage to nuclear power plants. This would enable the reactor to remain at nearly constant output, while cycling the electrical generator in response to the variability of the net load. This paper conceptually explores combinations of wind, solar, and nuclear that can provide a large fraction of a system's electricity, assuming the use of thermalenergystorage that would allow nuclear power to provide load following and cycling duty while operating at a constant reactor power output.}, journal = {Energy Policy}, author = {Denholm, P. and King, J.C. and Kutcher, C.F. and WIlson, P.P.H.}, month = may, year = {2012}, keywords = {Decarbonization, Electricity, Energy, Energy Storage, Nuclear, Power Plants, Renewable, Solar, Thermal, Wind}, pages = {301--311}, } T.D. Bohm, M.E. Sawan, S.T. Jackson, P.P.H. Wilson, "Detailed nuclear analysis of ITER ELM coils", Fusion Engineering and Design, , pp. (Feb 2012) Abstract BibTeX The ELM coils in ITER are intended to provide control of Edge Localized Modes (ELMs). These coils are located on the outboard side of ITER between the shield modules and vacuum vessel (VV) and are subject to high radiation levels. Detailed three-dimensional (3-D) models of the toroidal and poloidal legs of the ELM coil and surrounding region for the MCNP code were updated to reflect the latest design changes. Neutronics calculations were performed to determine a variety of radiation damage parameters for the ELM coils as well as the VV located behind them. Additionally, detailed CAD based models for the upper ELM coil region were used to perform a CAD based analysis using the DAG-MCNP5 code. The results show that the ELM coil will meet the specified material radiation limits. However, the nuclear heating on the vacuum vessel behind the poloidal multi-pipe manifolds will exceed the specified limit. @article{bohm_detailed_2012, title = {Detailed nuclear analysis of {ITER} {ELM} coils}, url = {http://www.sciencedirect.com/science/article/pii/S0920379612000427}, abstract = {The ELM coils in ITER are intended to provide control of Edge Localized Modes (ELMs). These coils are located on the outboard side of ITER between the shield modules and vacuum vessel (VV) and are subject to high radiation levels. Detailed three-dimensional (3-D) models of the toroidal and poloidal legs of the ELM coil and surrounding region for the MCNP code were updated to reflect the latest design changes. Neutronics calculations were performed to determine a variety of radiation damage parameters for the ELM coils as well as the VV located behind them. Additionally, detailed CAD based models for the upper ELM coil region were used to perform a CAD based analysis using the DAG-MCNP5 code. The results show that the ELM coil will meet the specified material radiation limits. However, the nuclear heating on the vacuum vessel behind the poloidal multi-pipe manifolds will exceed the specified limit.}, journal = {Fusion Engineering and Design}, author = {Bohm, T.D. and Sawan, M.E. and Jackson, S.T. and Wilson, P.P.H.}, month = feb, year = {2012}, keywords = {CAD, CAD, Coils, Edge Localized Modes, International Thermonuclear Reactor (ITER), MCNP, MCNP, Neutronics, Neutronics}, } J.A. Roberts, B.T. Rearden, P.P.H. Wilson, "Determination and Application of Partial Biases in Criticality Safety Validation", Nuclear Science and Engineering, , pp. (2012) BibTeX @article{roberts_determination_2012, title = {Determination and {Application} of {Partial} {Biases} in {Criticality} {Safety} {Validation}}, journal = {Nuclear Science and Engineering}, author = {Roberts, J.A. and Rearden, B.T. and Wilson, P.P.H.}, year = {2012}, } A. Ibrahim, M. Sawan, S.W. Mosher, T.M. Evans, D.E. Peplow, et al, "Global Evaluation of Prompt Dose Rates in ITER Using Hybrid Monte Carlo/Deterministic Techniques", Fusion Science and Technology, 60, pp. 676-680 (Aug 2011) Abstract BibTeX The hybrid Monte Carlo (MC)/deterministic techniques - Consistent Adjoint Driven Importance Sampling (CADIS) and Forward Weighted CADIS (FW-CADIS) - enable the full 3-D modeling of very large and complicated geometries. The ability of performing global MC calculations for nuclear parameters throughout the entire ITER reactor was demonstrated. The 2 m biological shield (bioshield) reduces the total prompt operational dose by six orders of magnitude. The divertor cryo-pump port results in a peaking factor of 120 in the prompt operational dose rate behind the bioshield of ITER. The equatorial port, plugged by 2 m of shielding, increases the prompt dose rate behind the bioshield by a factor of 47. The peak values of the prompt dose rates at the back surface of the bioshield were 240 Sv/hr and 94 Sv/hr corresponding to the regions behind the divertor cryo-pump port and the equatorial port, respectively. @article{ibrahim_global_2011, title = {Global {Evaluation} of {Prompt} {Dose} {Rates} in {ITER} {Using} {Hybrid} {Monte} {Carlo}/{Deterministic} {Techniques}}, volume = {60}, abstract = {The hybrid Monte Carlo (MC)/deterministic techniques - Consistent Adjoint Driven Importance Sampling (CADIS) and Forward Weighted CADIS (FW-CADIS) - enable the full 3-D modeling of very large and complicated geometries. The ability of performing global MC calculations for nuclear parameters throughout the entire ITER reactor was demonstrated. The 2 m biological shield (bioshield) reduces the total prompt operational dose by six orders of magnitude. The divertor cryo-pump port results in a peaking factor of 120 in the prompt operational dose rate behind the bioshield of ITER. The equatorial port, plugged by 2 m of shielding, increases the prompt dose rate behind the bioshield by a factor of 47. The peak values of the prompt dose rates at the back surface of the bioshield were 240 Sv/hr and 94 Sv/hr corresponding to the regions behind the divertor cryo-pump port and the equatorial port, respectively.}, number = {2}, journal = {Fusion Science and Technology}, author = {Ibrahim, A. and Sawan, M. and Mosher, S.W. and Evans, T.M. and Peplow, D.E. and Wilson, P. and Wagner, J.C.}, month = aug, year = {2011}, keywords = {Consistent Adjoint Driven Importance Sampling (CADIS), Deterministic, Dose Rates, Forward Weighted CADIS (FW-CADIS), Hybrid, International Thermonuclear Reactor (ITER), Monte Carlo Methods}, pages = {676--680}, } T.D. Bohm, B. Smith, M.E. Sawan, P.P.H. Wilson, "Assessment of the Surface Source Approach in 3-D Fusion Neutronics Analysis", Fusion Science and Technology, 60, pp. 703-707 (Aug 2011) Abstract BibTeX The surface source write/read capability in the 3-D neutronics code MCNP has been implemented in the CAD based DAG-MCNP. We performed neutronics calculations for a detailed solid model of an ITER first wall/shield module to assess the accuracy of the results obtained using the surface source for toroidal fusion systems. To further understand the sensitivity of the results to the size of the surface source and boundary conditions, we performed calculations for a simplified 3-D ITER model. The results show that use of the surface source approach is accurate provided that the surface source and associated reflective boundaries are extended beyond the component of interest by at least 10 cm and the surface source is generated/placed as close as possible to the front surface of that component. @article{bohm_assessment_2011, title = {Assessment of the {Surface} {Source} {Approach} in 3-{D} {Fusion} {Neutronics} {Analysis}}, volume = {60}, url = {http://fti.neep.wisc.edu/pdf/fdm1368.pdf}, abstract = {The surface source write/read capability in the 3-D neutronics code MCNP has been implemented in the CAD based DAG-MCNP. We performed neutronics calculations for a detailed solid model of an ITER first wall/shield module to assess the accuracy of the results obtained using the surface source for toroidal fusion systems. To further understand the sensitivity of the results to the size of the surface source and boundary conditions, we performed calculations for a simplified 3-D ITER model. The results show that use of the surface source approach is accurate provided that the surface source and associated reflective boundaries are extended beyond the component of interest by at least 10 cm and the surface source is generated/placed as close as possible to the front surface of that component.}, number = {2}, journal = {Fusion Science and Technology}, author = {Bohm, T.D. and Smith, B. and Sawan, M.E. and Wilson, P.P.H.}, month = aug, year = {2011}, keywords = {3D, Boundary Conditions, CAD, Code, International Thermonuclear Reactor (ITER), MCNP, Neutronics, Surface Source}, pages = {703--707}, } T.D. Bohm, M.E. Sawan, B. Smith, P.P.H. Wilson, "Investigation of Observed Peaking in Nuclear Paramters at Steel/Water Interfaces", Fusion Science and Technology, 60, pp. 698-702 (Aug 2011) Abstract BibTeX The ITER blanket modules (BM) are geometrically complex with many water coolant channels in a SS316 structure. Detailed mapping of nuclear heating, radiation damage, and helium production is an essential input to the design process. Previous high fidelity, high-resolution results calculated with the CAD based DAG-MCNP code revealed important heterogeneity effects on nuclear heating and helium production near steel/water interfaces. We carried out additional analysis for a simplified geometry to understand the reasons behind the observed peaking in the steel nuclear parameters at the interface with the water coolant. The results show that the peaking in nuclear heating is due to the softer neutron spectrum in the portion of steel adjacent to water which results in more gamma generation. Helium production peaking in steel adjacent to the water is due to the softer neutron spectrum which results in increased helium production primarily in B-10 impurities present in the SS316 in addition to a two-step reaction of low-energy neutrons with Ni. @article{bohm_investigation_2011, title = {Investigation of {Observed} {Peaking} in {Nuclear} {Paramters} at {Steel}/{Water} {Interfaces}}, volume = {60}, issn = {1536-1055}, abstract = {The ITER blanket modules (BM) are geometrically complex with many water coolant channels in a SS316 structure. Detailed mapping of nuclear heating, radiation damage, and helium production is an essential input to the design process. Previous high fidelity, high-resolution results calculated with the CAD based DAG-MCNP code revealed important heterogeneity effects on nuclear heating and helium production near steel/water interfaces. We carried out additional analysis for a simplified geometry to understand the reasons behind the observed peaking in the steel nuclear parameters at the interface with the water coolant. The results show that the peaking in nuclear heating is due to the softer neutron spectrum in the portion of steel adjacent to water which results in more gamma generation. Helium production peaking in steel adjacent to the water is due to the softer neutron spectrum which results in increased helium production primarily in B-10 impurities present in the SS316 in addition to a two-step reaction of low-energy neutrons with Ni.}, number = {2}, journal = {Fusion Science and Technology}, author = {Bohm, T.D. and Sawan, M.E. and Smith, B. and Wilson, P.P.H.}, month = aug, year = {2011}, keywords = {Blanket, Helium, Interfaces, International Thermonuclear Reactor (ITER), Nuclear Heating, Parameter, Steel, Water}, pages = {698--702}, } M.E. Sawan, A.M. Ibrahim, P.P.H. Wilson, E.P. Marriott, R.D. Stambaugh, et al, "Neutronics Analysis in Support of the Fusion Development Facility Design Evolution", Fusion Science and Technology, 60, pp. 671-675 (Aug 2011) Abstract BibTeX 3-D neutronics analysis was performed for the baseline design of FDF. Two blanket concepts were considered; Dual Coolant Lead Lithium (DCLL), and Helium Cooled Ceramic Breeder (HCCB). A peak outboard neutron wall loading of 2 MW/m2 and a fluence of 6 MW-yr/m2 can be achieved with 240 MW fusion power. The tritium breeding ratio is adequate for both blankets. Modest magnet damage parameters were obtained. However, it is recommended that the PF coil in the divertor region be moved vertically farther from the mid-plane to allow adding ~15 cm of shield to reduce the peak organic insulator dose to an acceptable level. @article{sawan_neutronics_2011, title = {Neutronics {Analysis} in {Support} of the {Fusion} {Development} {Facility} {Design} {Evolution}}, volume = {60}, abstract = {3-D neutronics analysis was performed for the baseline design of FDF. Two blanket concepts were considered; Dual Coolant Lead Lithium (DCLL), and Helium Cooled Ceramic Breeder (HCCB). A peak outboard neutron wall loading of 2 MW/m2 and a fluence of 6 MW-yr/m2 can be achieved with 240 MW fusion power. The tritium breeding ratio is adequate for both blankets. Modest magnet damage parameters were obtained. However, it is recommended that the PF coil in the divertor region be moved vertically farther from the mid-plane to allow adding {\textasciitilde}15 cm of shield to reduce the peak organic insulator dose to an acceptable level.}, number = {2}, journal = {Fusion Science and Technology}, author = {Sawan, M.E. and Ibrahim, A.M. and Wilson, P.P.H. and Marriott, E.P. and Stambaugh, R.D. and Wong, C.P.C.}, month = aug, year = {2011}, pages = {671--675}, } T.D. Bohm, S.T. Jackson, M.E. Sawan, P.P.H. Wilson, "Benchmarking a CAD-Based Monte Carlo Code Using Fusion-Specific Experiments", Nuclear Technology, 175, pp. 264-270 (July 2011) Abstract BibTeX Researchers at the University of Wisconsin-Madison Fusion Technology Institute and Argonne National Laboratories have recently developed a computer-aided-design-based Monte Carlo code (DAG-MCNP5) to perform nuclear analysis of complex three-dimensional systems such as ITER. In this work, DAG-MCNP5-calculated results will be compared to native MCNP5-calculated results and to measured results for ITER-specific benchmark experiments in order to provide additional quality assurance for DAG-MCNP. Calculated results are compared for the bulk shield mock-up and the helium-cooled pebble bed (HCPB) breeder blanket mock-up, which utilize the 14-MeV Frascati Neutron Generator facility. Neutron flux was measured at different depths in these experimental mock-ups using activation foils that cover the neutron energy range of 0 to 14 MeV. Additionally, tritium production in Li2CO3 pellets was measured in the HCPB experiment. Results of the foil activation calculations for the bulk shielding experiment and the HCPB breeder experiment show agreement within statistical uncertainty for DAG-MCNP5 and native MCNP5. Calculated results for tritium production in the HCPB mock-up also agree within statistical uncertainty for the DAG-MCNP5 and native MCNP5 calculations. Timing results showed that DAG-MCNP5 is 5.3 times slower than native MCNP5 for the bulk shield mock-up. For the HCPB mock-up, DAG-MCNP5 is 4.8 times slower than native MCNP5. It is concluded that the close agreement of calculated foil activation and tritium production between DAG-MCNP5 and native MCNP5 in these complex and ITER-relevant geometries provides additional quality assurance for the DAG-MCNP5 code and the mcnp2cad tool used in this work. @article{bohm_benchmarking_2011, title = {Benchmarking a {CAD}-{Based} {Monte} {Carlo} {Code} {Using} {Fusion}-{Specific} {Experiments}}, volume = {175}, abstract = {Researchers at the University of Wisconsin-Madison Fusion Technology Institute and Argonne National Laboratories have recently developed a computer-aided-design-based Monte Carlo code (DAG-MCNP5) to perform nuclear analysis of complex three-dimensional systems such as ITER. In this work, DAG-MCNP5-calculated results will be compared to native MCNP5-calculated results and to measured results for ITER-specific benchmark experiments in order to provide additional quality assurance for DAG-MCNP. Calculated results are compared for the bulk shield mock-up and the helium-cooled pebble bed (HCPB) breeder blanket mock-up, which utilize the 14-MeV Frascati Neutron Generator facility. Neutron flux was measured at different depths in these experimental mock-ups using activation foils that cover the neutron energy range of 0 to 14 MeV. Additionally, tritium production in Li2CO3 pellets was measured in the HCPB experiment. Results of the foil activation calculations for the bulk shielding experiment and the HCPB breeder experiment show agreement within statistical uncertainty for DAG-MCNP5 and native MCNP5. Calculated results for tritium production in the HCPB mock-up also agree within statistical uncertainty for the DAG-MCNP5 and native MCNP5 calculations. Timing results showed that DAG-MCNP5 is 5.3 times slower than native MCNP5 for the bulk shield mock-up. For the HCPB mock-up, DAG-MCNP5 is 4.8 times slower than native MCNP5. It is concluded that the close agreement of calculated foil activation and tritium production between DAG-MCNP5 and native MCNP5 in these complex and ITER-relevant geometries provides additional quality assurance for the DAG-MCNP5 code and the mcnp2cad tool used in this work.}, number = {1}, journal = {Nuclear Technology}, author = {Bohm, T.D. and Jackson, S.T. and Sawan, M.E. and Wilson, P.P.H.}, month = jul, year = {2011}, keywords = {Analysis, Benchmark, CAD, Code, DAG-MCNP5, Fusion, Helium, Helium-cooled Pebble Bed (HCPB), International Thermonuclear Reactor (ITER), Monte Carlo Methods}, pages = {264--270}, } Ahmad M. Ibrahim, Scott W. Mosher, Thomas M. Evans, Douglas E. Peplow, Mohamed E. Sawan, et al, "ITER Neutronics Modeling Using Hybrid Monte Carlo/Deterministic and CAD-Based Monte Carlo Methods", Nuclear Technology, 175, pp. 251-258 (Jul 2011) BibTeX @article{ibrahim_iter_2011, title = {{ITER} {Neutronics} {Modeling} {Using} {Hybrid} {Monte} {Carlo}/{Deterministic} and {CAD}-{Based} {Monte} {Carlo} {Methods}}, volume = {175}, journal = {Nuclear Technology}, author = {Ibrahim, Ahmad M. and Mosher, Scott W. and Evans, Thomas M. and Peplow, Douglas E. and Sawan, Mohamed E. and Wilson, Paul P.H. and Wagner, John C. and Heltemes, Thad}, month = jul, year = {2011}, pages = {251--258}, } B.C. Kiedrowski, F.B. Brown, P.P.H. Wilson, "Adjoint-Weighted Tallies for k-Eigenvalue Calculations with Continuous-Energy Monte Carlo", Nuclear Science and Engineering, 168, pp. 226-241 (Jul 2011) Abstract BibTeX A Monte Carlo method is developed that performs adjoint-weighted tallies in continuous-energy k-eigenvalue calculations. Each contribution to a tally score is weighted by an estimate of the relative magnitude of the fundamental adjoint mode, by way of the iterated fission probability, at the phase-space location of the contribution. The method is designed around the power iteration method such that no additional random walks are necessary, resulting in a minimal increase in computational time. The method is implemented in the Monte Carlo N-Particle (MCNP) code. These adjoint-weighted tallies are used to calculate adjoint-weighted fluxes, point reactor kinetics parameters, and reactivity changes from first-order perturbation theory. The results are benchmarked against discrete ordinates calculations, experimental measurements, and direct Monte Carlo calculations. @article{kiedrowski_adjoint-weighted_2011, title = {Adjoint-{Weighted} {Tallies} for k-{Eigenvalue} {Calculations} with {Continuous}-{Energy} {Monte} {Carlo}}, volume = {168}, abstract = {A Monte Carlo method is developed that performs adjoint-weighted tallies in continuous-energy k-eigenvalue calculations. Each contribution to a tally score is weighted by an estimate of the relative magnitude of the fundamental adjoint mode, by way of the iterated fission probability, at the phase-space location of the contribution. The method is designed around the power iteration method such that no additional random walks are necessary, resulting in a minimal increase in computational time. The method is implemented in the Monte Carlo N-Particle (MCNP) code. These adjoint-weighted tallies are used to calculate adjoint-weighted fluxes, point reactor kinetics parameters, and reactivity changes from first-order perturbation theory. The results are benchmarked against discrete ordinates calculations, experimental measurements, and direct Monte Carlo calculations.}, number = {3}, journal = {Nuclear Science and Engineering}, author = {Kiedrowski, B.C. and Brown, F.B. and Wilson, P.P.H.}, month = jul, year = {2011}, keywords = {Energy, Kinetics, MCNP, Monte Carlo Methods, k-Eigenvalue}, pages = {226--241}, } A.M. Ibrahim, S.W. Mosher, T.M. Evans, D.E. Peplow, M.E. Sawan, et al, "ITER Neutronics Modeling Using Hybrid Monte Carlo/Deterministic and CAD-based Monte Carlo Methods", Nuclear Technology, 175, pp. 251-258 (Jul 2011) Abstract BibTeX The immense size and complex geometry of the ITER experimental fusion reactor require the development of special techniques that can accurately and efficiently perform neutronics simulations with minimal human effort. This paper shows the effect of the hybrid Monte Carlo (MC)/deterministic techniques - Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) - in enhancing the efficiency of the neutronics modeling of ITER and demonstrates the applicability of coupling these methods with computer-aided-design-based MC. Three quantities were calculated in this analysis: the total nuclear heating in the inboard leg of the toroidal field coils (TFCs), the prompt dose outside the biological shield, and the total neutron and gamma fluxes over a mesh tally covering the entire reactor. The use of FW-CADIS in estimating the nuclear heating in the inboard TFCs resulted in a factor of [approximately]275 increase in the MC figure of merit (FOM) compared with analog MC and a factor of [approximately]9 compared with the traditional methods of variance reduction. By providing a factor of [approximately]21 000 increase in the MC FOM, the radiation dose calculation showed how the CADIS method can be effectively used in the simulation of problems that are practically impossible using analog MC. The total flux calculation demonstrated the ability of FW-CADIS to simultaneously enhance the MC statistical precision throughout the entire ITER geometry. Collectively, these calculations demonstrate the ability of the hybrid techniques to accurately model very challenging shielding problems in reasonable execution times. @article{ibrahim_iter_2011, title = {{ITER} {Neutronics} {Modeling} {Using} {Hybrid} {Monte} {Carlo}/{Deterministic} and {CAD}-based {Monte} {Carlo} {Methods}}, volume = {175}, url = {http://www.ornl.gov/sci/scale/pubs/RPSD_2010_Ibrahim.pdf}, abstract = {The immense size and complex geometry of the ITER experimental fusion reactor require the development of special techniques that can accurately and efficiently perform neutronics simulations with minimal human effort. This paper shows the effect of the hybrid Monte Carlo (MC)/deterministic techniques - Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) - in enhancing the efficiency of the neutronics modeling of ITER and demonstrates the applicability of coupling these methods with computer-aided-design-based MC. Three quantities were calculated in this analysis: the total nuclear heating in the inboard leg of the toroidal field coils (TFCs), the prompt dose outside the biological shield, and the total neutron and gamma fluxes over a mesh tally covering the entire reactor. The use of FW-CADIS in estimating the nuclear heating in the inboard TFCs resulted in a factor of [approximately]275 increase in the MC figure of merit (FOM) compared with analog MC and a factor of [approximately]9 compared with the traditional methods of variance reduction. By providing a factor of [approximately]21 000 increase in the MC FOM, the radiation dose calculation showed how the CADIS method can be effectively used in the simulation of problems that are practically impossible using analog MC. The total flux calculation demonstrated the ability of FW-CADIS to simultaneously enhance the MC statistical precision throughout the entire ITER geometry. Collectively, these calculations demonstrate the ability of the hybrid techniques to accurately model very challenging shielding problems in reasonable execution times.}, number = {1}, journal = {Nuclear Technology}, author = {Ibrahim, A.M. and Mosher, S.W. and Evans, T.M. and Peplow, D.E. and Sawan, M.E. and Wilson, P.P.H. and Wagner, J.C.}, month = jul, year = {2011}, keywords = {CAD, Consistent Adjoint Driven Importance Sampling (CADIS), Deterministic, Forward Weighted CADIS (FW-CADIS), Fusion reactors, Hybrid, International Thermonuclear Reactor (ITER), Modeling, Monte Carlo Methods, Neutronics}, pages = {251--258}, }