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DOE OSTI2018

Generation of the Cross Section Library for PROTEUS

The cross sections for a high-fidelity neutron transport code PROTEUS can be generated in multiple ways. Multigroup cross sections in the ISOTXS format can be generated using the external codes such as MC 2 -3 and Monte Carlo codes such as Serpent and OpenMC. For the Monte Carlo codes, their multigroup cross section outputs should be converted to the ISOTXS format using a post processor named GenISOTXS developed by ANL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

(U) LNK3DNT Capability in SENSMG

The SENSMG multigroup neutron sensitivity code has been extended to read materials and geometry from redoin and lnk3dnt files written by the PARTISN multigroup discrete ordinates code or the MCNP6 Monte Carlo code. This report describes the user interface for this capability and the code verification that was done. Not all of the capability of SENSMG is available for redoin/lnk3dnt inputs. The only responses allowed are k eff or α eigenvalues. Fixed-source problems are not allowed, and there is no mechanism to input reaction rates or reaction-rate ratios. The next section of this report describes the user interface. Section III describes the verification of the capability. Section IV is a summary and conclusions. The input files used for verification are listed in the appendix.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2019

(U) Adjoint-Weighted Leakage Time in SENSMG

It is desired to compute the adjoint weighted leakage time $τ_1$ for an α-eigenvalue problem. The reciprocal of the leakage time is $\frac{1}{τ_1}$ = $\frac{\langle ψ^*, Ω∙∇ψ \rangle}{\langle ψ^*, \frac{1}{v}ψ \rangle}$ where $ψ^*$ and $ψ$ are the adjoint and forward α-eigenfunctions, respectively; $v$ is neutron speed; and the angle brackets indicate an inner product over all independent variables, in this case space, energy, and angle. This memo discusses the implementation of Eq. (1) in the multigroup neutron sensitivity code SENSMG (Refs. 2 and 3), which relies for the neutron transport on the multigroup discrete ordinates code PARTISN (Ref. 4).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

(U) LNK3DNT Capability in SENSMG Extended

The SENSMG multigroup neutron sensitivity code can read materials and geometry from redoin and lnk3dnt files written by the PARTISN multigroup discrete ordinates code or the MCNP6 Monte Carlo code. This report describes the user interface for this capability and the code verification that was done. Reference 5 described an initial version of this capability. Two important extensions have been made: 1) Reaction-rate ratios are now allowed as responses with redoin/lnk3dnt inputs; and 2) Sensitivities of reaction-rate ratios to mass densities and interface locations are now calculated (for all problems, not just redoin/lnk3dnt problems).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

(U) Transport Corrections in SENSMG

Truncating the spherical harmonics expansion of the neutron scattering source in the Boltzmann transport equation leads to truncation error. Methods for approximately correcting for this truncation error are called transport corrections . The PARTISN multigroup discrete ordinates neutron transport code has several transport correction options. These options can be used in the SENSMG multigroup neutron sensitivity code, which relies on PARTISN for the neutron transport. However, with one exception, the sensitivities computed by SENSMG do not include the effect of transport corrections. Thus, when certain transport corrections are used, SENSMG results are inconsistent with the transport solutions. This report discusses the issues surrounding computing sensitivities when transport corrections are used. Problems and possible solutions are identified.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

(U) Transport Corrections Implemented in SENSMG

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

(U) LNK3DNT Capability in SENSMG Tested with Transport Correction

The SENSMG multigroup neutron sensitivity code can read materials and geometry from redoin and lnk3dnt files written by the PARTISN multigroup discrete ordinates code or the MCNP6 Monte Carlo code. Reference 5 described an initial version of this capability. Reference 6 introduced two important extensions: 1) Reaction-rate ratios are now allowed as responses with redoin/lnk3dnt inputs; and 2) Sensitivities of reaction-rate ratios to mass densities and interface locations are now calculated (for all problems, not just redoin/lnk3dnt problems). In this report, we document the testing of the redoin/lnk3dnt capability when used with transport correction. PARTISN’s transport corrections are described in Ref. 3, and Refs. 7 and 8 discuss the calculation of cross-section sensitivities when transport corrections are applied. References 7 and 8 verify SENSMG for problems with transport correction using central differences. This report verifies the capability with redoin/lnk3dnt inputs by comparison with regular SENSMG inputs. Unlike Refs. 5 and 6, in this report we use a spherical test problem because the sensitivity of the response with respect to the highest-order scattering is zero in cylinders due to the use of flux moments in the inner products.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2021

Recent Updates in ETOE-2 and MC 2 -3

The MC 2 -3 code is a multigroup cross section generation code for fast reactor applications, developed by Argonne National Laboratory (ANL) under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. In this work, first, the cross section libraries for ENDF/B-VII.1 and ENDF/B-VIII.0 data were generated using the updated ETOE-2 code. Preliminary verification and validation tests of the ENDF/B-VII.1 MC 2 -3 library were performed with the selected fast reactor benchmark problems. Processing and verifying the libraries suggested that significant efforts would be required for thoroughly verifying the ENDF/B-VII.1 MC 2 -3 library and successfully processing the ENDF/B-VIII.0 MC 2 -3 library. Secondly, the cross section generation capability of MC 2 -3 was updated with the intermediate group lattice calculation and the equivalent Dancoff-factor cell (EDC) method to significantly improve the performance of a twodimensional assembly calculation using the method of characteristics (MOC). This effort was made useful in implementing and verifying the EDC method in the Griffin cross section API. Finally, we analyzed a SFR problem, for which unusually large deviations in core eigenvalues from Monte Carlo solutions were reported, by generating multigroup cross sections with MC 2 -3, performing core calculations with DIF3D-VARIANT, and analyzing cross section and eigenvalue results against Monte Carlo solutions. The analysis confirmed that the MC 2 -3/DIF3D solutions were in good agreement with Monte Carlo solutions, providing the correct process of accurately generating broad-group cross sections with MC 2 -3

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2022

Transport @ LANL: An Overview [Slides]

Deterministic neutron transport requires discretization of the Boltzmann transport equation in space, angle, energy, and time. Discretization in energy is typically performed with the multigroup approximation, which tries to preserve continuous energy cross section information by using a weighted average over discrete energy groups. There are two important requirements for accurate multigroup cross sections: appropriate energy group boundaries and an accurate weight function. The former can be selected with appropriate optimizers; the latter either requires flux solutions from comparable simulations or expert knowledge.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2022

Modeling, Performance Assessment, and Nodal Data Analysis of TRISO-Fueled Systems with Shift

This technical report documents several enhancements to the Shift Monte Carlo (MC) code under the US Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in fiscal year (FY) 2022. Performance enhancements were added to Shift specifically for tristructural isotropic (TRISO)–fueled reactor systems and guided based on performance analysis in FY 2021. For the pebble performance model developed in previous studies, the runtime improved by ~ 91× compared to the original model and ~ 2× compared to the user-optimized model. Compared to Serpent, Shift is ~ 3× slower if Serpent delta-tracking is enabled but ~ 2× faster when delta-tracking is disabled. The multigroup cross section generation was improved through simplifying tally input definitions, porting several post-processing tally operations from Python scripts into the Shift code base, and accounting for production reactions in the scattering multiplicity. Progress was also made on two emerging capabilities: (1) the development of Titan (a Shift reactor physics user interface) and (2) initial investigation into path-length tallies for computing multigroup scattering matrices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Integration of Online Cross-Section Generation Capability with Depletion and Transient Solvers in Griffin

Griffin is a Multiphysics Object-Oriented Simulation Environment (MOOSE)-based reactor multiphysics analysis application jointly developed by Argonne and Idaho National Laboratories under the DOENE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. In FY25, an online crosssection generation capability based on the Self-Shielding Application Programming Interface (SSAPI) was demonstrated for TRISO-fueled reactor problems under steady-state conditions. This fiscal year, that capability was extended to support depletion and transient multiphysics calculations, enabling high-fidelity analyses that generate self-shielded cross sections on the fly from the actual evolving composition and temperature states rather than from pre-tabulated libraries. For depletion, a two-way coupling was established in which SSAPI computes compact-averaged self-shielded cross sections that the depletion solver then uses to advance the Bateman equations, with the updated compositions returned to SSAPI at each step; the depletion module was refactored to support both library-based and SSAPI-based cross sections, and additional logic was added to track daughter isotopes and to exclude minor isotopes for efficiency. For transient analysis, the SSAPI multigroup library was extended with the kinetics data required for time-dependent calculations, the Improved Quasi-Static (IQS) scheme was coupled with SSAPI, and several supporting capabilities were implemented, including a self-shielding treatment that lets control rods and drums move within a self-shielded model, which had previously been impossible and had ruled out rod- and drum-movement transients with on-the-fly cross sections altogether, a new mixing scheme for delayed-neutron precursor decay constants, a checkpoint-based restart workflow, and performance improvements such as pointwise cross-section interpolation and the bypassing of unnecessary Dancoff factor calculations. The implemented capabilities were verified against Serpent Monte Carlo solutions. For depletion, a prismatic pin-cell problem based on a Next Generation Nuclear Plant (NGNP) Very High Temperature Reactor benchmark showed excellent agreement, with eigenvalue differences within 200 pcm over the entire burnup range (up to 140 MWD/kgU) and fission-product and actinide inventories agreeing to within 0.8% and 2.5%, respectively; a heat-pipe microreactor assembly problem with a much higher fuel loading confirmed the same behavior and quantified the bias introduced when the multigroup equivalence effect is neglected. For transient analysis, a pin-cell problem with a step reactivity insertion and temperature feedback reproduced the analytically expected asymptotic power and showed close agreement between the direct and IQS solutions, and a two-dimensional microreactor core problem with control-drum rotation exercised the new moving-drum self-shielding treatment and demonstrated successful coupling of the online crosssection generation with both the direct and IQS transient methods. The capability was further exercised on a full-core pebble-bed problem, in which Griffin was coupled with the System Analysis Module (SAM) to simulate load-following operation of the gPBR with the Doppler feedback resolved at the TRISO fuel kernel temperature. These developments in Griffin provide a convenient, high-fidelity approach to cross-section generation for advanced thermal reactors with geometrically complex and highly heterogeneous configurations, including TRISO-fueled prismatic and pebble-bed systems, and support steady-state, depletion, and transient multiphysics calculations. They also enable self-shielded cross sections to be evaluated directly at the actual coupled state of the system, thereby establishing a foundation for high-fidelity, fully coupled multiphysics analysis of advanced reactors

Park, H.↗

DOE OSTI2025

Barriers and Benefits: Understanding Riders’ Views on Pooled Rideshare in the U.S.

This manuscript provides actionable recommendations to enhance user satisfaction and address existing barriers regarding pooled rideshare (PR) in the United States. Despite PR’s intended benefits, such as reduced traffic congestion and cost savings, its adoption remains limited. To identify these actionable items, a U.S. nationwide survey with 5385 participants explored transportation preferences, barriers, and motivators for PR use in the summer of 2021. First, two factor analyses were conducted. The first factor analysis identified the five factors associated with one’s willingness to consider PR (time/cost, traffic/environment, safety, privacy, and service experience). The second factor analysis revealed the four factors related to ways to optimize one’s PR experience (comfort/ease of use, convenience, vehicle technology/accessibility, and passenger safety). Privacy concerns, for instance, were found to reduce the likelihood of PR adoption by 77%, and convenience had the potential to increase it by 156%. A structural equation model evaluated the relationships among these nine key factors influencing PR usage to develop the Pooled Rideshare Acceptance Model (PRAM). The privacy, safety, trust service, and convenience factors each had a significant large effect (Cohen’s f 2 > 0.35) on the model. PRAM was extended using multigroup analyses to reveal the nuanced impact of 16 demographics, including gender, generation, rideshare experience, etc., highlighting the need for tailored strategies to improve PR acceptance through the Pooled Rideshare Acceptance Model Multigroup Analyses (PRAMMAs). Multiple workshops were held with diverse audiences to translate the team’s findings to date into 84 actionable recommendations, categorized across topical areas like safety, routing, driver and passenger selection, user education, etc. These findings are a foundation for a future study to determine which items resonate with different user groups. In the meantime, the actional items serve as a user-driven resource for policymakers, transportation network companies, and researchers, offering a roadmap to potential improvements to PR services to address existing concerns with the goal of increasing the usage of PR.

actionable recommendations↗

DOE OSTI2017

Very Low-energy Supernovae: Light Curves and Spectra of Shock Breakout

The brief transient emitted as a shock wave erupts through the surface of a pre-supernova star carrying information about the stellar radius and explosion energy. Here, the CASTRO code, which treats radiation transport using multigroup flux-limited diffusion, is used to simulate the light curves and spectra of shock breakout in very low-energy supernovae (VLE SNe), explosions in giant stars with final kinetic energy much less than 10{sup 51} erg. VLE SN light curves, computed here with the KEPLER code, are distinctively faint, red, and long-lived, making them challenging to find with transient surveys. The accompanying shock breakouts are brighter, though briefer, and potentially easier to detect. Previous analytic work provides general guidance, but numerical simulations are challenging, due to the range of conditions and lack of equilibration between color and effective temperatures. We consider previous analytic work and extend discussions of color temperature and opacity to the lower energy range explored by these events. Since this is the first application of the CASTRO code to shock breakout, test simulations of normal energy shock breakout of SN 1987A are carried out and compared with the literature. A set of breakout light curves and spectra are then calculated for VLE SNe with final kinetic energies in the range 10{sup 47}--10{sup 50} erg for red supergiants with main-sequence masses of 15 and 25 M{sub ⊙}. The importance of uncertainties in stellar atmosphere model, opacity, and ambient medium is discussed, as are observational prospects with current and forthcoming missions.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2017

Light Curves and Spectra from a Unimodal Core-collapse Supernova

To assess the effectiveness of optical emission as a probe of spatial asymmetry in core-collapse supernovae (CCSNe), we apply in this paper the radiative transfer software SuperNu to a unimodal CCSN model. The SNSPH radiation hydrodynamics software was used to simulate an asymmetric explosion of a 16 ${M}_{\odot }$ zero-age main-sequence binary star. The ejecta has 3.36 ${M}_{\odot }$ with 0.024 ${M}_{\odot }$ of radioactive 56 Ni, with unipolar asymmetry along the z-axis. For 96 discrete angular views, we find a ratio between maximum and minimum peak total luminosities of ~1.36. The brightest light curves emerge from views orthogonal to the z-axis. Multigroup spectra from UV to IR are obtained. We find a shift in wavelength with viewing angle in a near-IR Ca ii emission feature, consistent with Ca being mostly in the unimode. We compare emission from the gray gamma-ray transfer in SuperNu and from the detailed gamma-ray transfer code Maverick. Relative to the optical light curves, the brightness of the gamma-ray emission is more monotonic with respect to viewing angle. UBVRI broadband light curves are also calculated. Parallel with the unimode, the U and B bands have excess luminosity at $\gtrsim 10$ days post-explosion, due to 56 Ni on the unimode. Finally, we compare our CCSN model with SN 2002ap, which is thought to have a similar ejecta morphology.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2019

Fornax: A Flexible Code for Multiphysics Astrophysical Simulations

This paper describes the design and implementation of our new multigroup, multidimensional radiation hydrodynamics code Fornax and provides a suite of code tests to validate its application in a wide range of physical regimes. Instead of focusing exclusively on tests of neutrino radiation hydrodynamics relevant to the core-collapse supernova problem for which Fornax is primarily intended, we present here classical and rigorous demonstrations of code performance relevant to a broad range of multidimensional hydrodynamic and multigroup radiation hydrodynamic problems. Our code solves the comoving-frame radiation moment equations using the M1 closure, utilizes conservative high-order reconstruction, employs semi-explicit matter and radiation transport via a high-order time stepping scheme, and is suitable for application to a wide range of astrophysical problems. To this end, we first describe the philosophy, algorithms, and methodologies of Fornax and then perform numerous stringent code tests that collectively and vigorously exercise the code, demonstrate the excellent numerical fidelity with which it captures the many physical effects of radiation hydrodynamics, and show excellent strong scaling well above 100,000 MPI tasks.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2011

Sensitivities and Uncertainties Related to Numerics and Building Features in Urban Modeling

Oak Ridge National Laboratory (ORNL) has been engaged in the development and testing of a computational system that would use a grid of activation foil detectors to provide postdetonation forensic information from a nuclear device detonation. ORNL has developed a high-performance, three-dimensional (3-D) deterministic radiation transport code called Denovo. Denovo solves the multigroup discrete ordinates (SN) equations and can output 3-D data in a platform-independent format that can be efficiently analyzed using parallel, high-performance visualization tools. To evaluate the sensitivities and uncertainties associated with the deterministic computational method numerics, a numerical study on the New York City Times Square model was conducted using Denovo. In particular, the sensitivities and uncertainties associated with various components of the calculational method were systematically investigated, including (a) the Legendre polynomial expansion order of the scattering cross sections, (b) the angular quadrature, (c) multigroup energy binning, (d) spatial mesh sizes, (e) the material compositions of the building models, (f) the composition of the foundations upon which the buildings rest (e.g., ground, concrete, or asphalt), and (g) the amount of detail included in the building models. Although Denovo may calculate the idealized model well, there may be uncertainty in the results because of slight departures of the above-named parameters from those used in the idealized calculations. Fluxes and activities at selected locations from perturbed calculations are compared with corresponding values from the idealized or base case to determine the sensitivities associated with specified parameter changes. Results indicate that uncertainties related to numerics can be controlled by using higher fidelity models, but more work is needed to control the uncertainties related to the model.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2013

Validation of Criticality Safety Calculations with SCALE 6.2

SCALE 6.2 provides numerous updates in nuclear data, nuclear data processing, and computational tools utilized in the criticality safety calculational sequences relative to SCALE 6.1. A new 252-group ENDF/B-VII.0 multigroup neutron library, improved ENDF/B-VII.0 continuous energy data, as well as the previously deployed 238-group ENDF/B-VII.0 neutron library are included in SCALE 6.2 for criticality safety analysis. The performance of all three libraries for keff calculations is examined with a broad sampling of critical experiment models covering a range of fuels and moderators. Critical experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) that are available in the SCALE Verified, Archived Library of Inputs and Data (VALID) are used in this validation effort. Over 300 cases are used in the validation of KENO V.a, and a more limited set of approximately 50 configurations are used for KENO-VI validation. Additionally, some KENO V.a cases are converted to KENO-VI models so that an equivalent set of experiments can be used to validate both codes. For continuous-energy calculations, SCALE 6.2 provides improved performance relative to SCALE 6.1 in most areas with notable improvements in fuel pin lattice cases, particularly those with mixed oxide fuel. Multigroup calculations with the 252-group library also demonstrate improved performance for fuel lattices, uranium (high and intermediate enrichment) and plutonium metal experiments, and plutonium solution systems. Overall, SCALE 6.2 provides equivalent or smaller biases than SCALE 6.1, and the two versions of KENO provide similar results on the same suite of problems.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

Development of a SCALE Tool for Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

Two methods for calculating eigenvalue sensitivity coefficients in continuous-energy Monte Carlo applications were implemented in the KENO code within the SCALE code package. The methods were used to calculate sensitivity coefficients for several criticality safety problems and produced sensitivity coefficients that agreed well with both reference sensitivities and multigroup TSUNAMI-3D sensitivity coefficients. The newly developed CLUTCH method was observed to produce sensitivity coefficients with high figures of merit and low memory requirements, and both continuous-energy sensitivity methods met or exceeded the accuracy of the multigroup TSUNAMI-3D calculations.

CLUTCH↗