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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Space-Dependent Wielandt Shifts for Multigroup Diffusion Eigenvalue Problems

We present a new concept—the space-dependent Wielandt shift (SDWS)—for accelerating the convergence of the power iteration (PI) scheme for multigroup diffusion k-eigenvalue problems. The SDWS improves on standard Wielandt shift (WS) techniques, which are empirical in nature and are typically effective only when the current estimate of the solution is reasonably converged. By accounting for the physics of the problem through SDWS, we are able to improve the acceleration for the initial iterates when the current estimate of the solution is not close to convergence. Numerical results from one-dimensional problems suggest that, compared to standard WS techniques, the new SDWS techniques can provide upward of a 46% reduction in the number of PIs required for convergence and a 40% reduction in the computational time required. This improvement is sensitive to several problem-dependent factors, such as the geometry and energy-dependence of the problem, the spatial discretization, and the initial guess. The reduction in computational time is also dependent on the linear solver in the PI scheme, as it is well known that WSs can significantly worsen the conditioning of the diffusion linear system. In this paper, we provide a detailed study of the impact of WSs on the performance of several iterative linear solvers. Results from our implementation of SDWS in the three-dimensional (3D) code MPACT show that SDWS can provide similar speedups for 3D multigroup diffusion eigenvalue problems. These results also show that moderate speedups can be obtained by applying SDWS to the coarse mesh finite difference (CMFD) solver in a CMFD-accelerated transport scheme. However, the benefit of doing this may be limited because all but the first few CMFD solves are relatively easy to converge, regardless of the WS used.

Nuclear Science & Technology↗
DOE OSTI2019

The Subray Method of Characteristics

The MPACT code is being jointly developed by the University of Michigan and Oak Ridge National Laboratory. It uses the 2-D/1-D method to solve neutron transport problems for reactors. The 2-D/1-D method decomposes the problem into a stack of 2-D planes and uses a high-fidelity transport method to resolve all heterogeneity in each plane. These planes are then coupled axially, using a lower-order solver. With this scheme, three-dimensional (3-D) solutions to the transport equation can be obtained at a much lower cost. The 2-D/1-D method assumes that the materials are axially homogeneous for each 2-D plane. Violation of this assumption requires homogenization, which can significantly reduce the accuracy of the calculation. This work proposes the subray method of characteristics (subray MOC) as a solution to this problem. Subray MOC is a subgrid method that allows local heterogeneities to be directly resolved by method of characteristics while treating the rest of the 2-D plane as axially uniform. This improves the accuracy in the neighborhood of the heterogeneity while minimizing the increase in run time. The method was applied to variations of the C5G7 benchmark problems and compared with a previously developed subgrid method called the subplane collision probabilities (SCP) method. Comparisons were made among results obtained using subray MOC, the SCP method, and no subgrid method. Subray MOC consistently performed best, reducing maximum 3-D power distribution errors from as high as 30% to 2% or less. Moreover, it consistently outperformed the SCP method with run times that were shorter than the reference calculations.

2-D/1-D method↗
DOE OSTI2014

How gravitational lensing helps γ-ray photons avoid γ – γ absorption

We investigate potential γ – γ absorption of γ-ray emission from blazars arising from inhomogeneities along the line of sight, beyond the diffuse Extragalactic Background Light (EBL). As plausible sources of excess γ – γ opacity, we consider (1) foreground galaxies, including cases in which this configuration leads to strong gravitational lensing, (2) individual stars within these foreground galaxies, and (3) individual stars within our own galaxy, which may act as lenses for microlensing events. We found that intervening galaxies close to the line of sight are unlikely to lead to significant excess γ – γ absorption. This opens up the prospect of detecting lensed gamma-ray blazars at energies above 10 GeV with their gamma-ray spectra effectively only affected by the EBL. The most luminous stars located either in intervening galaxies or in our galaxy provide an environment in which these gamma-rays could, in principle, be significantly absorbed. However, despite a large microlensing probability due to stars located in intervening galaxies, γ-rays avoid absorption by being deflected by the gravitational potentials of such intervening stars to projected distances ({sup i}mpact parameters{sup )} where the resulting γ – γ opacities are negligible. Thus, neither of the intervening excess photon fields considered here, provide a substantial source of excess γ – γ opacity beyond the EBL, even in the case of very close alignments between the background blazar and a foreground star or galaxy.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2022

Analytic error analysis of cross section interpolation methods in nodal diffusion codes - II: Numerical results

This paper is the second part of a two-part paper that documents the numerical results for the partial derivatives model presented in part I. In this paper, we derive the error bounds for the analytical point-wise error expression and verify our bounds with numerical experiments. The point-wise error expressions make available, and bound, the sources that contribute to the total error of the interpolated cross section in terms of the Lagrange interpolation errors and the model form error. MPACT is used to generate two-group homogenized cross sections for Westinghouse's AP1000 Region 4 lattice to evaluate the accuracy of the bounds. Error bounds calculated over a grid are compared to numerical data for uni-variate and multi-variate interpolation. The point-wise error bounds of a typical case matrix - two branches in each state variable - are displayed for bi-variate interpolation in the state variables: moderator density, fuel temperature, and boron concentration. The error bounds are shown to be highly accurate compared to numerical results, and in accordance with the underlying physics. We then discuss and show how the sources of error contribute to the total error, and consider the improvement of each error source. Finally, we mention future work such as propagating our cross section error bounds through a reactivity calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Lead Slowing Down Spectrometer Status Report

This report documents the progress that has been completed in the first half of FY2012 in the MPACT-funded Lead Slowing Down Spectrometer project. Significant progress has been made on the algorithm development. We have an improve understanding of the experimental responses in LSDS for fuel-related material. The calibration of the ultra-depleted uranium foils was completed, but the results are inconsistent from measurement to measurement. Future work includes developing a conceptual model of an LSDS system to assay plutonium in used fuel, improving agreement between simulations and measurement, design of a thorium fission chamber, and evaluation of additional detector techniques.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Lead Slowing Down Spectrometer Research Plans

The MPACT-funded Lead Slowing Down Spectrometry (LSDS) project has been evaluating the feasibility of using LSDS techniques to assay fissile isotopes in used nuclear fuel assemblies. The approach has the potential to provide considerable improvement in the assay of fissile isotopic masses in fuel assemblies compared to other non-destructive techniques in a direct and independent manner. The LSDS collaborations suggests that the next step to in empirically testing the feasibility is to conduct measurements on fresh fuel assemblies to understand investigate self-attenuation and fresh mixed-oxide (MOX) fuel rodlets so we may better understand extraction of masses for 235 U and 239 Pu. While progressing toward these goals, the collaboration also strongly suggests the continued development of enabling technology such as detector development and algorithm development, which could provide significant performance benefits.

Lead slowing down spectrometry↗
DOE OSTI2013

A New 2D-Transport, 1D-Diffusion Approximation of the Boltzmann Transport equation

The work performed in this project consisted of the derivation, implementation, and testing of a new, computationally advantageous approximation to the 3D Boltzmann transport equation. The solution of the Boltzmann equation is the neutron flux in nuclear reactor cores and shields, but solving this equation is difficult and costly. The new “2D/1D” approximation takes advantage of a special geometric feature of typical 3D reactors to approximate the neutron transport physics in a specific (ax- ial) direction, but not in the other two (radial) directions. The resulting equation is much less expensive to solve computationally, and its solutions are expected to be sufficiently accurate for many practical problems. In this project we formulated the new equation, discretized it using standard methods, developed a stable iteration scheme for solving the equation, implemented the new numerical scheme in the MPACT code, and tested the method on several realistic problems. All the hoped- for features of this new approximation were seen. For large, difficult problems, the resulting 2D/1D solution is highly accurate, and is calculated about 100 times faster than a 3D discrete ordinates simulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

A Deployable Fast-Neutron Coded-Aperture Imager for Quantifying Nuclear Material

This brief report describes a newly constructed fast-neutron coded-aperture imager intended for performing quantitative measurements of plutonium holdup at fuel cycle facilities. The present report was prepared in fulfillment of milestone M4FT-13OR0402053, “Construct Imager,” of Fuel Cycle R&D Material Protection, Accounting, and Control Technology (MPACT) work package FT-13OR040205, “Fast-Neutron Imaging to Quantify Nuclear Materials - ORNL.” The first measurements taken with the imager will be reported later this year. For quantifying plutonium holdup, fast-neutron imaging is an appealing mode of measurement in scenarios where contact measurements with non-imaging detectors may be impractical or their results ambiguous because of surrounding material or nearby sources. The use of fast neutrons (as opposed to the more numerous gamma rays) is appealing because of their ability to penetrate equipment and nuclear material that is opaque to gamma radiation. However, the ability of fast neutrons to penetrate material also leads to challenges in constructing a fast-neutron imager sufficiently compact to be amenable for use in a facility while remaining sufficiently sensitive to neutron emanations from plutonium or other actinides. Furthermore, while fast-neutron detection is commonly achieved via organic liquid scintillators, use of the imager in a production environment favors mechanically robust detectors that do not contain flammable liquid. In response to these challenges, Oak Ridge National Laboratory (ORNL) has built a fast-neutron imager that has a compact footprint and whose segmented neutron detectors are built from the first cast plate of the recently developed plastic scintillator with pulse shape discrimination properties developed by Lawrence Livermore National Laboratory and commercially available from Eljen Technologies as experimental scintillator EJ-299-34.

Hausladen, Paul↗
DOE OSTI2013

The Deployable Fast-Neutron Coded-Aperture Imager: Demonstration of Locating One or More Sources in Three Dimensions

This brief report describes commissioning measurements performed using a newly constructed fast neutron coded-aperture imager intended for performing quantitative measurements of plutonium holdup at fuel cycle facilities. The present report was prepared in fulfillment of milestone M3FT-13OR0402054, “Imager commissioning and 3D source configuration measurements” of Fuel Cycle R&D Material Protection, Accounting, and Control Technology (MPACT) work package FT-13OR040205, “Fast Neutron Imaging to Quantify Nuclear Materials - ORNL.” In addition to representing the first measurements made with the new device, these measurements also demonstrate for the first time the ability to localize one or more neutron sources in three dimensions by combining imaging data from multiple measurements in which the imager was positioned at different locations. The development and use of fast-neutron imaging technology are motivated by a desire to locate and quantify plutonium holdup in scenarios where contact measurements with non-imaging detectors may be impractical or their results ambiguous because of surrounding material or nearby sources. Considering the goal of quantitative measurements, fast neutrons (as opposed to the more numerous gamma rays) are preferred in many situations because of their ability to penetrate equipment and nuclear material that are essentially opaque to gamma radiation. However, there are many challenges associated with constructing a practical imager for fast neutrons. For example, the ability of fast neutrons to penetrate material also leads to challenges in constructing a fast-neutron imager that is sufficiently compact to be suitable for use in a facility while remaining sufficiently sensitive to neutron emanations from plutonium or other actinides. Moreover, use of the imager in a production environment favors mechanically robust detectors that do not contain flammable liquid, unlike the organic liquid scintillators that are typically used to achieve fast-neutron detection. Last of all, a single-sided measurement geometry, as is typically used in imaging applications, necessitates accurate measurement of the location of a deposit of plutonium holdup in three dimensions in order to correctly infer the source strength. In response to these challenges, Oak Ridge National Laboratory (ORNL) has built a fast-neutron imager that has a compact footprint and whose segmented neutron detectors are built from the first cast plate of the recently developed plastic scintillator with pulse shape discrimination properties developed by Lawrence Livermore National Laboratory and commercially available from Eljen Technologies as experimental scintillator EJ-299-34. This report presents a description and preliminary analysis of mock holdup measurements intended to determine the three-dimensional location of one or more neutron sources placed in prototypical equipment from a set of imaging measurements that scanned the extent of the equipment.

Hausladen, Paul↗
DOE OSTI2016

SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation

The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Extended Dry Storage Signature Bench Scale Detector Conceptual Design

This report is the conceptual design of a detector based on research within the Extended Dry Storage Signature Development project under the DOE-­NE MPACT campaign. This is the second year of the project; from this year’s positive results, the next step is building a prototype and testing with real materials .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

MOC Efficiency Improvements Using a Jacobi Inscatter Approximation

In recent weeks, attention has been given to resolving the convergence issues encountered with TCP 0 by trying a Jacobi (J) inscatter approach when group sweeping, where the inscatter source is constructed using the previous iteration flux. This is in contrast to a Gauss-Seidel (GS) approach, which has been the default to-date, where the scattering source uses the most up-to-date flux values. The former is consistent with CASMO, which has no issues with TCP 0 convergence. Testing this out on a variety of problems has demonstrated that the Jacobi approach does indeed provide substantially more stability, though can take more outer iterations to converge. While this is not surprising, there are improvements that can be made to the MOC sweeper to capitalize on the Jacobi approximation and provide substantial speedup. For example, the loop over groups, which has traditionally been the outermost loop in MPACT, can be moved to the interior, avoiding duplicate modular ray trace and coarse ray trace setup (mapping coarse mesh surface indexes), which needs to be performed repeatedly when group is outermost.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Voltammetry Method Evaluation

The purpose of the ANL MPACT Voltammetry project is to evaluate the suitability of previously developed cyclic voltammetry techniques to provide electroanalytical measurements of actinide concentrations in realistic used fuel processing scenarios. The molten salts in these scenarios are very challenging as they include high concentrations of multiple electrochemically active species, thereby creating a variety of complications. Some of the problems that arise therein include issues related to uncompensated resistance, cylindrical diffusion, and alloying of the electrodeposited metals. Improvements to the existing voltammetry technique to account for these issues have been implemented, resulting in good measurements of actinide concentrations across a wide range of adverse conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Signatures of Extended Storage of Used Nuclear Fuel in Casks

As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Test Plan for Cask Identification Detector

This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench-scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Performance Improvements for Coarse Mesh Finite Difference Acceleration L3:RTM.PRT.P13.02

The development of VERA-CS in recent years has focused on developing the capability to simulate multiple cycles of operating commercial nuclear power plants. Now that these capabilities have advanced to the point where it is being deployed to users, the focus is on improving the computational performance of various components in VERA-CS. In this work, the focus is on the Coarse Mesh Finite Difference (CMFD) solution in MPACT. CMFD serves multiple purposes in the 2D/1D solution methodology. First, it is a natural mechanism to tie together the radial MOC transport and the axial SP3 solution. Because the CMFD system solves the multigroup three-dimensional core in one system, it pulls together the global response of the system. In addition, the CMFD solution provides a framework to accelerate the convergence of the eigenvalue problem.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Standalone BISON Through VERA: User's Manual

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA Core Simulator (VERA-CS) development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code being developed by the Idaho National Laboratory (INL) is the focus of ongoing work in the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework, BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗