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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

A new multigroup method for cross-sections that vary rapidly in energy

Here, we present a numerical method for solving the time-independent thermal radiative transfer (TRT) equation or the neutron transport (NT) equation when the opacity (cross-section) varies rapidly in frequency (energy) on the microscale ε; ε corresponds to the characteristic spacing between absorption lines or resonances, and is much smaller than the macroscopic frequency (energy) variation of interest. The approach is based on a rigorous homogenization of the TRT/NT equation in the frequency (energy) variable. Discretization of the homogenized TRT/NT equation results in a multigroup-type system, and can therefore be solved by standard methods.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Feasibility of a multigroup Boltzmann–Fokker–Planck solution for electron beam dose calculations

Legacy nuclear-reactor Boltzmann solvers start clinical deployment as an alternative to Monte Carlo (MC) codes and Fermi–Eyges semiemprical models in radiation oncology treatment planning. Today’s certified clinical solvers are limited to photon beams. In this paper, ELECTR, a state-of-the-art multigroup electron cross sections generation module in NJOY is presented and validated against Lockwood’s calorimetric measurements, EGS-nrc and GEANT-4 for 1–20 MeV unidirectional electron beams. The nuclear-reactor DRAGON-5 solver is upgraded to access the library and solve the Boltzmann–Fokker–Planck (BFP) equation. A variety of heterogeneous radiotherapy and radiosurgery phantom configurations were used for validation purpose. Case studies include a thorax benchmark, that of a typical breast Intra-Operative Radiotherapy and a high-heterogeneity patient-like benchmark. For all beams, 100% of the water voxels satisfied the American Association of Physicists in Medicine accuracy criterion for a BFP-MC dose error below 2%. At least, 97.0% of adipose, muscle, bone, lung, tumor and breast voxels satisfied the 2% criterion. The average BFP-MC relative error was about 0.56% for all voxels, beams and materials combined. By irradiating homogeneous slabs from Z = 1 (hydrogen) to Z = 99 (einsteinium), we reported performance and defects of the CEPXS mode [US. Sandia National Lab., SAND-89-1685] in ELECTR for the entire periodic table. For all Lockwood’s benchmarks, NJOY-DRAGON dose predictions are within the experimental data precision for 98% of voxels.

42 ENGINEERING↗
DOE OSTI2023

The Random Ray Method Versus Multigroup Monte Carlo: The Method of Characteristics in OpenMC and SCONE

The Random Ray Method (TRRM) is a recently developed approach to solving neutral particle transport problems based on the Method of Characteristics. While the method previously has been implemented only in closed-source or limited-functionality codes, this work describes its implementation in two open-source Monte Carlo codes: OpenMC and SCONE. The random ray implementations required small modifications to the existing Multigroup Monte Carlo (MGMC) solvers, offering a rare venue for redundant, fine-grained, "apples-to-apples" speed and accuracy comparisons between transport methods. To this end, TRRM and MGMC solvers are evaluated against each other using each code's native capabilities on reactor eigenvalue problems with different degrees of energy discretization. On the C5G7 benchmark (featuring only seven energy groups), TRRM achieves a maximum pin power error comparable to or lower than that of MGMC for a given run time. On a problem with 69 energy groups, MGMC is found to scale more efficiently, obtaining a lower pin power error for a given run time. However, the defining difference between the two transport methods is found to be their vastly different uncertainty distributions. Specifically, TRRM is found to maintain similar levels of accuracy and uncertainty throughout the simulation domain whereas MGMC can exhibit orders-of-magnitude greater errors in areas of the problem that feature low neutron flux. For instance, TRRM provided an up to 373 times speed advantage compared with MGMC for computing the flux in low-flux regions in the moderator surrounding the C5G7 core.

42 ENGINEERING↗
DOE OSTI2020

Stretched and Filtered Multigroup Pn Transport for Improved Positivity and Accuracy

One of the common methods for solving the radiation transport equation is to use a polynomial expansion for the angle variable(s). Recent research that reduces the oscillations and improves the positivity of the gray transport equation solutions is here applied to the multigroup transport equations. Constant scale factors that stretch the time axis and constant scattering opacities that filter the solution greatly increase the accuracy of the solution with no added nonlinearities. No new solution techniques are required. Test problems are presented in one and two dimensions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Adapting CLUTCH methodology to multigroup TSUNAMI-3D for eigenvalue sensitivity calculations

The sensitivity of the eigenvalue to uncertainties in nuclear data and its evaluation are important for nuclear criticality safety. TSUNAMI-3D sequences within the SCALE code system offer several options to the user community for calculating eigenvalue sensitivity coefficients with multigroup (MG) and continuous energy (CE) 3D transport capabilities. TSUNAMI-3D sequences implement the adjoint-based perturbation theory with MG KENO code, the Contributon Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance CHaracterization (CLUTCH) method with CE KENO code, and the Iterated Fission Probability (IFP) method with CE KENO and Shift codes. Each method has benefits and limitations depending on the problem that is run. The work presented here aims to adapt the CLUTCH method, which enables the Contributon method's mesh-free, memory-efficient approach for calculating adjoint-weighted tallies for sensitivity calculations, to the MG TSUNAMI-3D sequence. This application would eliminate the explicit adjoint KENO calculation, as well as the memory-consuming mesh flux moment tallies required by the conventional MG TSUNAMI-3D. Smaller memory footprints in the CLUTCH methodology and relatively shorter runtimes in MG KENO transport can make MG TSUNAMI-3D a viable method for some complex problems. Moreover, this adaptation allows MG sensitivity calculations with Shift, ORNL's next-generation high-performance Monte Carlo transport code, which currently does not offer any sensitivity capabilities with MG particle transport simulations. Initial implementation of the new MG TSUNAMI-3D sequence and its preliminary results with a selected critical benchmark experiment in the Verified, Archived Library of Inputs and Data (VALID) are presented in this study.

KENO↗
DOE OSTI2012

Cross-section uncertainty propagation in the multigroup slowing-down equations using probability table formalism

In the frame of neutral-particle (neutron, gamma) transport, the uncertainty propagation calculation regarding the uncertainties on cross sections is often carried out without explicitly taking into account their probabilistic distribution. We investigate a new uncertainty propagation formalism where the cross-section uncertainty distributions are represented by probability tables. This technical note develops this approach for the steady-state slowing-down equation without up-scattering and in an infinite medium. This work is based on a deterministic multiband formalism that takes into account multilevel probability tables for cross sections. The first level represents the variation of cross sections versus lethargy (or energy) in each group of the multigroup lethargy mesh and thus corresponds to the classical cross-section probability tables. The higher levels represent the uncertainties on each step of the first-level cross-section probability table. This method is validated against a Monte Carlo calculation in a case of neutron slowing down in a {sup 238}U-hydrogen homogeneous mixture, showing fully consistent numerical results. The main interest of the deterministic multilevel multiband formalism is that it gives not only the mean value and the variance but also a probabilistic distribution of the fluxes. In the near future, we plan to investigate more deeply the robustness of this new approach in relation to high values of cross-section uncertainties and to introduce cross-section uncertainty correlations as well. Meanwhile, the promise of this work is its extension to the general transport steady-state equation solved by the discrete ordinates (S(N)) or Monte Carlo methods. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Multigroup Radiation-Hydrodynamics with a High-Order, Low-Order Method

Recent efforts at Los Alamos National Laboratory to develop a moment-based, scale-bridging [or high-order (HO)–low-order (LO)] algorithm for solving large varieties of the transport (kinetic) systems have shown promising results. A part of our ongoing effort is incorporating this methodology into the framework of the Eulerian Applications Project to achieve algorithmic acceleration of radiationhydrodynamics simulations in production software. By starting from the thermal radiative transfer equations with a simple material-motion correction, we derive a discretely consistent energy balance equation (LO equation). We demonstrate that the corresponding LO system for the Monte Carlo HO solver is closely related to the original LO system without material-motion corrections. We test the implementation on a radiative shock problem and show consistency between the energy densities and temperatures in the HO and LO solutions as well as agreement with the semianalytic solution. We also test the approach on a more challenging two-dimensional problem and demonstrate accuracy enhancements and algorithmic speedups. This paper extends a recent conference paper by including multigroup effects.

42 ENGINEERING↗
DOE OSTI2012

NDI Multigroup Radiochemistry Dosimetry Library hoffman2009

We have processed the modeled neutron induced nuclear reaction cross section data from Hoffman et al. into NDI 618 lanl multigroup neutron dosimetry tables for yttrium, zirconium, and rhodium, and created corresponding radiochemistry chain files. The resulting cross sections for 87g Y(n,γ) 88g Y and 88g Y(n,2n) 87g Y compare favorably to the data in the rc2004 618 library and are similar to the data in the crsllnl library. We have released these data and chains to facilitate the direct comparison of yttrium ratios for Los Alamos and Livermore data and request feedback from our user community on the usefulness of this data before we process more data.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very high void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1998

A Multigroup Method for the Calculation of Neutron Fluence with a Source Term

Current research on the Grant involves the development of a multigroup method for the calculation of low energy evaporation neutron fluences associated with the Boltzmann equation. This research will enable one to predict radiation exposure under a variety of circumstances. Knowledge of radiation exposure in a free-space environment is a necessity for space travel, high altitude space planes and satellite design. This is because certain radiation environments can cause damage to biological and electronic systems involving both short term and long term effects. By having apriori knowledge of the environment one can use prediction techniques to estimate radiation damage to such systems. Appropriate shielding can be designed to protect both humans and electronic systems that are exposed to a known radiation environment. This is the goal of the current research efforts involving the multi-group method and the Green's function approach.

Heinbockel, J. H.↗
DOE OSTI2012

The Method of Characteristics for 2-D Multigroup and Pointwise Transport Calculations in SCALE/CENTRM

SCALE 6 computes problem-dependent multigroup (MG) cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic pointwise (PW) transport calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The PW calculation is performed by the CENTRM code using a 1-D cylindrical Wigner-Seitz model with the white boundary condition instead of the real rectangular cell shape to represent a lattice unit cell. The pointwise fluxes computed by CENTRM are not exact because a 1-D model is used for the transport calculation, which introduces discrepancies in the MG self-shielded cross sections, resulting in some deviation in the eigenvalue. In order to solve this problem, the method of characteristics (MOC) has been applied to enable the CENTRM PW transport calculation for a 2-D square pin cell. The computation results show that the new BONAMI/CENTRM-MOC procedure produces very precise self-shielded cross sections compared to MCNP reaction rates.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Improvement of SCALE-XSPROC Multigroup Cross Section Processing Based on the CENTRM Pointwise Slowing Down Calculation

The SCALE-XSProc multigroup (MG) cross section processing procedure based on the CENTRM pointwise slowing down calculation is the primary procedure to process problem-dependent self-shielded MG cross sections and scattering matrices for neutron transport calculations.This procedure supports various cell-based geometries including slab, 1-D cylindrical, 1-D spherical and 2-D rectangular configurations and doubly heterogeneous particulate fuels. Recently, this procedure has been significantly improved to be applied to any advanced reactor analysis covering thermal and fast reactor systems, and to be comparable to continuous energy (CE) Monte Carlo calculations. Some reactivity bias and reaction rate differences have been observed compared with CE Monte Carlo calculations, and several areas for improvement have been identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range, (2) 10 eV thermal cut-off energy for the free gas model, (3) on-the-fly adjustments to the thermal scattering matrix, (4) normalization of the pointwise neutron flux, and (5) fine MG energy structure. This procedure ensures very accurate MG cross section processing for high-fidelity deterministic reactor physics analysis for various advanced reactor systems.

Kim, Kang Seog↗
DOE OSTI2012

Improvements in transport calculations by the optimized multigroup libraries for fast neutron systems

This paper shows how to improve the accuracy of the transport calculations using in the APOLLO2 code the optimized multigroup libraries calculated by AEMC for fast neutron systems. These ameliorations concern the fission source calculation and the self-shielding models. The calculation of the fission source was generalized to fission spectra including an incident neutron energy dependence. The subgroup self-shielding model was updated for a mixture of resonant nuclides. Some tests on a Pu-239 sphere without reflectors and a fast sodium cell show that the use of four fission spectra guarantees a correct representation of the fission source. The test on a Pu-239 sphere with a thick steel reflector proves that the subgroup self-shielding, accounting for the mutual shielding of several resonant nuclides, allows us to improve the accuracy of the neutron transport solution in the reflector. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Hybrid method of deterministic and probabilistic approaches for multigroup neutron transport problem

A hybrid method of deterministic and probabilistic methods is proposed to solve Boltzmann transport equation. The new method uses a deterministic method, Method of Characteristics (MOC), for the fast and thermal neutron energy ranges and a probabilistic method, Monte Carlo (MC), for the intermediate resonance energy range. The hybrid method, in case of continuous energy problem, will be able to take advantage of fast MOC calculation and accurate resonance self shielding treatment of MC method. As a proof of principle, this paper presents the hybrid methodology applied to a multigroup form of Boltzmann transport equation and confirms that the hybrid method can produce consistent results with MC and MOC methods. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗