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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 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Use of SCALE Continuous-Energy Monte Carlo Tools for Eigenvalue Sensitivity Coefficient Calculations

The TSUNAMI code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The CLUTCH and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE KENO framework to generate the capability for TSUNAMI-3D to perform eigenvalue sensitivity calculations in continuous-energy applications. This work explores the improvements in accuracy that can be gained in eigenvalue and eigenvalue sensitivity calculations through the use of the SCALE CE KENO and CE TSUNAMI continuous-energy Monte Carlo tools as compared to multigroup tools. The CE KENO and CE TSUNAMI tools were used to analyze two difficult models of critical benchmarks, and produced eigenvalue and eigenvalue sensitivity coefficient results that showed a marked improvement in accuracy. The CLUTCH sensitivity method in particular excelled in terms of efficiency and computational memory requirements.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

Enhancements in Continuous-Energy Monte Carlo Capabilities for SCALE 6.2

SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, industry, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a plug-and-play framework that includes three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 provides several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, sensitivity and uncertainty analysis, and improved fidelity in nuclear data libraries. A brief overview of SCALE capabilities is provided with emphasis on new features for SCALE 6.2.

Continuous-Energy↗
DOE OSTI2014

Sandia Computational Engine for Particle Transport for Radiation Effects v 1.4

The SCEPTRE code solves the linear Boltzmann transport equation for general two- and three-dimensional geometries. SCEPTRE is capable of handling any particle type for which multigroup-Legendre cross sections are available. However, the code is designed primarily to model the transport of photons, electrons, and positrons through matter.

Source record↗
DOE OSTI2016

NJOY2016

NJOY2016 is used to convert evaluations in the Evaluated Nuclear Data Files (ENDF) format into forms useful for practical applications such as fission and fusion reactor analysis, criticality safety, radiation shielding, nuclear waste management, nuclear medicine procedures, and more. It begins with the generation of pointwise libraries, including reaction and resonance reconstruction, Doppler broadening, radiation heating and damage, thermal scattering data, unresolved resonance data, and gas production. It reviews the production of libraries for the continuous-energy Monte Carlo code MCNP, multigroup neutron, photon, and particle cross sections and matrices, and photon interaction data. Uncertainty information for ENDF data is processed, including the capability for calculating covariances of resonance data, angular distributions, energy distributions, and radioactive nuclide production. NJOY can prepare and process thermal scattering data evaluations into application libraries. It can also plot cross sections, angular distributions and secondary emission spectra.

Kahler III, A. C. (Skip)↗
DOE OSTI2017

Branson: A Mini-App for Studying Parallel IMC, Version 1.0

This code solves the gray thermal radiative transfer (TRT) equations in parallel using simple opacities and Cartesian meshes. Although Branson solves the TRT equations it is not designed to model radiation transport: Branson contains simple physics and does not have a multigroup treatment, nor can it use physical material data. The opacities have are simple polynomials in temperature there is a limited ability to specify complex geometries and sources. Branson was designed only to capture the computational demands of production IMC codes, especially in large parallel runs. It was also intended to foster collaboration with vendors, universities and other DOE partners. Branson is similar in character to the neutron transport proxy-app Quicksilver from LLNL, which was recently open-sourced.

Long, Alex↗
DOE OSTI2017

Demonstration of MAMMOTH Strongly-Coupled Multiphysics Simulation with the Godiva Benchmark Problem

This paper provides a detailed description of the Godiva benchmark problem suitable for multiphysics code verification. Multigroup transport weak forms for general curvilinear coordinates are derived using SAAF (self-adjoint angular flux) formulation. A new scheme based on spherical harmonics expansion for evaluating the angular derivative terms using the SN method are presented and tested. A one-dimensional spherical coordinate system is used by MAMMOTH for modeling the Godiva benchmark. Strongly-coupled multiphysics simulation mode of MAMMOTH is exercised and numerical results are given.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

ENDF/B-VIII.0 Testing With AMPX and SCALE

As the new ENDF/B-VIII.0 data has continued development from its initial beta release into the final release version, it has subsequently undergone processing and testing at Oak Ridge National Laboratory with AMPX and SCALE. Each of the beta releases has been processed with AMPX to create both multigroup (MG) and continuous-energy (CE) libraries. The MG libraries were all processed using the SCALE 252-group structure. These libraries were then used with the VALID suite for benchmarking the new data library’s performance against the currently distributed ENDF/B-VII.1 (VII.1)-based SCALE libraries. Here, we compare the MG and CE VALID results for the two current ENDF releases, versions VII.1 and the most recent beta release of ENDF/B-VIII.0 as of this writing (VIII.0.β4). Generally, both the MG and CE VALID aggregate results for VIII.0.β4 compare reasonably well with their VII.1-derived counterparts. As can be expected of any new ENDF release, some of the individual cases have substantially different computed neutron multiplicity (k eff ) values between the two releases, although typically the differences are less than 100 percent-milli (pcm). However, changes to the plutonium cross sections have induced much larger changes in the ratio of the computed k eff to the experimental value (C/E) for thermal spectrum plutonium solution critical systems. This paper details the comparative performance of the VII.1 and VIII.0.β4 data for the broad range of test cases covered by the VALID suite. A companion paper at this conference provides details of the baseline validation of SCALE 6.2, including the first release of the VII.1 data.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

A Pebble Tracking Transport Algorithm for Pebble Bed Reactor Analysis

Pebble bed reactor (PBR) designs pose unique challenges to modeling and simulation. The challenges originate from the multiple levels of heterogeneity partly due to random packing in and flow through the core of fuel pebbles. This paper proposes a novel pebble tracking transport (PTT) algorithm to address some of the challenges facing PBR analysis. Each individual pebble is tracked and modeled on an unstructured mesh of which the nodes are aligned with the pebble centroids. The neutron transport equation is solved on this pebble tracking mesh with pre-assembled elemental matrices. These elemental matrices are independent of the cross sections and the streaming directions, thus can be assembled before the calculation without introducing too much memory overhead especially when combined with spatial domain decomposition. Numerical results show that these matrices can be assembled efficiently. PTT is implemented in Rattlesnake, the INL MOOSE-based multigroup radiation transport application. Both the angular and spatial convergence are investigated and the results indicate that low order angular and spatial resolution is typically sufficient for analysis. A complete work flow for PBR analysis with PTT including, pebble packing, meshing, cross section generation, solution of the transport equation, postprocessing, etc. is demonstrated with a simplified reactor with 41,048 pebbles.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Iterated Discontinuity Factors Based on Reference Outgoing Currents for Discontinuous Finite Element Diffusion

We present a novel formulation for iteratively computing Discontinuity Factors (DF) designed to reproduce reference partial outgoing currents at the interface between macro-regions using Discontinuous Galerkin Finite Elements to solve the multigroup neutron diffusion equation. Similarly, an expression for the boundary coefficients (BCf) is proposed to match the partial outgoing current at vacuum boundaries. Upon convergence, macro-region averaged reaction and leakage rates are preserved. In addition, we consider a source-free steady-state problem depending on the DFs. The existence of a non-trivial solution is due to the non-linearity of the problem. This form greatly simplifies the implementation within Rattlesnake and efficiently converges by updating the DFs and BCfs at every linear iteration of the Preconditioned Jacobian-Free Newton-Krylov solver. The main challenge then lies in determining an appropriate scaling factor to ensure that the initial values for the DFs and BCfs are not too far from their converged value. While this factor does not change the final solution, it has the potential to significantly affect convergence, for purely numerical reasons. An expression based on the reference fission source integral is proposed and satisfying results are presented for problems and compared to Superhomogénéisation (SPH) results with both reflecting and vacuum boundary conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Iterated Discontinuity Factors Based on Reference Outgoing Currents for Discontinuous Finite Element Diffusion

We present a novel formulation for iteratively computing Discontinuity Factors (DF) designed to reproduce reference partial outgoing currents at the interface between macro-regions using Discontinuous Galerkin Finite Elements to solve the multigroup neutron diffusion equation. Similarly, an expression for the boundary coefficients (BCf) is proposed to match the partial outgoing current at vacuum boundaries. Upon convergence, macro-region averaged reaction and leakage rates are preserved. In addition, we consider a source-free steady-state problem depending on the DFs. The existence of a non-trivial solution is due to the non-linearity of the problem. This form greatly simplifies the implementation within Rattlesnake and efficiently converges by updating the DFs and BCfs at every linear iteration of the Preconditioned Jacobian-Free Newton-Krylov solver. The main challenge then lies in determining an appropriate scaling factor to ensure that the initial values for the DFs and BCfs are not too far from their converged value. While this factor does not change the final solution, it has the potential to significantly affect convergence, for purely numerical reasons. An expression based on the reference fission source integral is proposed and satisfying results are presented for problems and compared to Superhomogénéisation (SPH) results with both reflecting and vacuum boundary conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

α-Weighted Transition Rate Matrix Method

A more direct approach in calculating the Transition Rate Matrix Method operators and the significance of this method are presented. Consequences of using α-spectrum as the Transition Rate Matrix Method weighting function instead of the κ-spectrum are investigated. A prompt supercritical infinite medium multigroup system is considered. Results on continuous problem using OpenMC will be provided in the full paper.

Variansyah, I.↗
DOE OSTI2019

Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗
DOE OSTI2019

VALIDATION OF KENO V.A AND KENO-VI IN SCALE 6.3 BETA 3 USING ENDF/B-VII.1 AND ENDF/B-VIII LIBRARIES

In early 2019, internal beta releases of SCALE 6.3 were identified for internal users, as well as a limited number of external users, to apply when testing key features. Recently developed AMPX-formatted libraries based on ENDF/B-VIII have also been generated for use in internal testing. In this effort, a criticality safety validation was performed using nuclear data libraries based on the ENDF/B-VII.1 and ENDF/B-VIII releases to quantify the changes caused by code development and to identify differences caused by the new nuclear data. As with past validations, the primary suite of experiments used is the Verified, Archived Library of Inputs and Data (VALID) maintained in the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The VALID suite contains a total of 618 individual configurations, including systems with fast, thermal, mixed, and intermediate neutron spectra. Fissile species include a range of uranium enrichments, plutonium, mixtures of uranium and plutonium, and 233U. Metal, solution, and compound, mostly in pin array fissile forms, are included in the suite. The results presented here include the 252-group general purpose multigroup library based on ENDF/B-VII.1 and the continuous-energy libraries based on ENDF/B-VII.1 and ENDF/B-VIII.

Marshall, William (B.J.)↗
DOE OSTI2019

Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-Group Libraries for Advanced Reactor Analysis

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.

Kim, Kang Seog↗
DOE OSTI2020

Multigrid Algorithms with Projection and Prolongation over Elements of the Phase Space for K-Eigenvalue Transport Problems

This paper describes new multilevel acceleration methods for solving the multigroup neu- tron transport eigenvalue problems. These multilevel algorithms use different projection and prolongation operators in the phase space. The Nonlinear Diffusion Acceleration (NDA) method with multiple grids in energy is formulated with the prolongation oper- ator based on multiplication iterative correction and linear-in-energy mapping. Another multilevel NDA method uses the projection operator with coarsening in energy between the high-order transport and low-order NDA equations. The third algorithm is formu- lated with the partial-current based CMFD low-order equations and applies projection operators in space and energy. The numerical results are presented.

Cornejo, Luke↗
DOE OSTI2022

Analysis of SCALE Criticality and Sensitivity Calculations for Reflected HEU Cylinders [Abstract]

The SCALE code package offers several nuclear data libraries to support Monte Carlo (MC) transport, as well as MC-based derivation of $\kappa$ eff sensitivity and uncertainty (S/U) data. The CSAS sequence using the KENO MC code can utilize continuous-energy (CE) cross sections, or pre-generated multigroup (MG) cross section libraries. The use of MG libraries introduces bias into calculations in exchange for faster transport solutions. The TSUNAMI-3D sequence also utilizes KENO MC calculations. TSUNAMI-3D has two CE calculational methods: the Iterated Fission Probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. Previous work has shown poor agreement between CLUTCH and confirmatory direct perturbation calculations in specific applications, e.g., fissionable and polyethylene reflectors

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications

A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

The fast neutron fluence and the activation detector activity calculations using the effective source method and the adjoint function

This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weighting is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with few group method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reacto rand located axially at the position of maximum power and at the position of the weld. Both of these methods (the effective source and the adjoint function) are briefly described in the present paper. The paper also describes their application to the solution of fast neutron fluence and detectors activities for the VVER-440 reactor. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗