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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Bison Improvements for Robustness and Speed

Early in its development, the primary emphasis of efforts to improve the Bison fuel performance code were focused on initial development of new capabilities. As Bison has matured and become adopted by a wider set of users, it has become clear that a number of robustness issues needed to be addressed to ensure that the code can reliably produce a converged solution. This need became particularly evident when a recent effort by Westinghouse Electric Company to simulate all of the fuel rods in a reactor using Bison resulted in approximately 30% of the models failing to run to completion because of lack of convergence or other errors that were encountered. It is important to emphasize that there was no evidence that Bison generated results that were incorrect for the set of models exercised. In many cases, issues such as those encountered could be addressed by adjusting solver-related parameters. However, there is no question that for Bison to become widely adopted for production work, it needs to be inherently far more robust than it was. Because of this, a concerted effort was undertaken to improve Bison's robustness, or its ability to reliably obtain converged solutions under a wider variety of conditions. As a result of this work, a number of high-priority areas for improvement for modeling light water reactor (LWR) fuel rods have been identified and addressed. As a result of this work, the number of failing Bison analyses in the aforementioned set of simulations for a full core has been reduced to a fraction of a percent. In addition, the robustness of Bison models using the VERA core simulator has been significantly improved. This is important because all of the Bison models representing individual fuel rods must run to completion successfully for the full core model to run to completion. This report provides summaries of the identified sources of robustness issues in Bison and the development that was done to resolve these issues. It also summarizes the addition of new capabilities to address deficiencies in Bison's ability to include needed aspects of LWR fuel behavior, and improvements that were made to improve Bison's solution efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig

Fuel rods in a pressurized water nuclear reactor (PWR) experience coolant flow-induced relative motion against the grid support features, a phenomenon called grid-to-rod fretting (GTRF), which may cause progressive fretting wear damage. Wear-through of the cladding on fuel rods will leak the radioactive fuel into the coolant loop to significantly increase the radiation level in PWR primary loop and often trigger expensive Post Irradiation Examination. In this work, a unique autoclave fretting test rig was designed and fabricated to allow studying GTRF using actual cladding and grid materials. Water temperature was up to 220 °C, which was designed to mimic the environment in an industry full-assembly reactor core simulator. The contact load (0.1–1 N), oscillation frequency (20–30 Hz) and stroke (50–150 µm), and work rate (1–3 mW) were defined based on simulations of GTRF in an actual reactor. Using this autoclave GTRF rig, tests were conducted to learn the effect of water temperature and to investigate the wear behavior of different cladding-grid material combinations currently used in actual PWRs, including Zr alloy claddings without and with pre-oxidation against Zr alloy and Inconel grids. Higher water temperature evidently increased the cladding wear, and pre-oxidation of the Zr alloy and/or using the Inconel grid effectively reduced the wear rate. In conclusion, results were correlated with the data of a full-size industry reactor core simulator in terms of both wear rate and wear scar morphology.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

SCALE 6.2 Lattice Physics Performance Assessment

The US Nuclear Regulatory Commission relies on the lattice physics analysis capabilities of the SCALE code system to perform confirmatory licensing analyses. Either SCALE lattice physics code—TRITON/NEWT or Polaris—can be used to generate cross section data used by the PARCS nodal core simulator for full-core neutronics calculations. This report presents an assessment of the accuracy of SCALE lattice physics codes for preparation of lattice physics data that are used to support simulator codes such as the NRC’s PARCS, for UO 2 -mixed oxide (MOX)/Zr fueled light water reactor (LWR) analyses. Due to the nature of lattice physics calculations, critical reactor experiment benchmarks cannot be modeled in explicit detail in a lattice physics code. However, this limitation does not mean that these measurement data are not usable for lattice physics studies. Therefore, either geometry approximations or axial buckling must be implemented to determine the critical water height. These modeling limitations have led to development of a three-phase assessment strategy. In the first phase, selected critical experiment benchmarks are modeled using the SCALE 3D continuous-energy (CE) Monte Carlo (MC) code KENO, which is the most rigorous neutron transport method available in SCALE, with no approximations in the spatial, angle, or energy treatments. Biases and statistical uncertainties in quantities of interest such as k eff and pin power distributions are determined by comparing CE KENO results to experimental data. This first phase in the accuracy assessment is to demonstrate and establish the use of CE KENO as a reference solution for the second phase. In the second phase, 14 numerical test suites are used to compare SCALE lattice physics calculations with CE KENO as a reference solution. In the third phase, SCALE lattice physics depletion calculations are performed, and the spent fuel isotopic results are compared with available radiochemical assay measurements. Isotopic measurement comparisons provide quantitative assessment of isotopic density distribution predictions with the depletion models in SCALE lattice physics codes. This report documents results for all test suites. The assessment was performed using standard production techniques unless otherwise noted. Both TRITON/NEWT and Polaris exhibited acceptable accuracy for most test cases. For the few test cases in which acceptable accuracy criteria were not met, further code and data development are planned. The computer codes used in this assessment are as follows: SCALE 6.2 rev19189 (pre-release of 6.2.1) was used for CE KENO, TRITON/NEWT and Polaris (PWR only) calculations; SCALE 6.2.2 was used for Polaris calculations to address several updates in support of support boiling water reactor (BWR) lattice geometries. For this work, there is no difference between SCALE 6.2.1 and SCALE 6.2.2 beyond the inclusion of the Polaris BWR analysis capability; The ENDF/B-VII.1 continuous energy and 252 group libraries deployed in SCALE 6.2 were generated with AMPX 6.2 that is distributed with SCALE.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Emergence of robust anthropogenic increase of heat stress-related variables projected from CORDEX-CORE climate simulations

The information of when and where region-specific patterns in both mean and extreme temperatures leading to heat stress will emerge from the present-day climate variability is important to plan adaptation options, but to date studies on this issue still remain limited and fragmented. Here, we estimate the time of emergence (TOE) of temperature and wet-bulb temperature (Tw), a better indication of heat stress, using fine-scale, long-term regional climate model projections under the RCP2.6 and RCP8.5 scenarios across six different domains. Differently from previous studies, the TOE is determined using three methods applied on impact-relevant variables: two different signal-to-noise frameworks based on summer mean temperature and Tw and a statistical test to identify significant differences in daily extreme distributions. The TOE response to RCP2.6 and RCP8.5 with respect to the end of 20th century variability differs significantly regardless of which TOE metric is applied. For summer mean temperature, the land fraction reaching TOE is expected to exceed 90% by the 2050s under the RCP8.5, whereas the increase rate of land exposure to TOE tends to stagnate over time under the RCP2.6 so that more than 40% of land will not experience TOE by the end of the 21st century. Compared to temperature, the TOE of Tw is reached earlier in most of the wet tropics but is delayed in hot and dry regions because of the nonlinear response of Tw to humidity. For both temperature and Tw, the TOE appears earlier in regions with low baseline variability, such as in the tropics. Despite the uncertainties arising from the choice of TOE metrics, the vast majority of regions in Africa and southeast Asia experience TOE in the early 21st century under both the RCP2.6 and RCP8.5 scenarios, which stresses the urgent need for developing adequate adaptation strategies in these regions.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Consistent coupling algorithms for coupled core-edge simulations of plasma turbulence

Two schemes for coupling gyrokinetic simulations of microturbulence in tokamaks are proposed. The first scheme is based on an additive Schwarz domain decomposition. In this work, we show that, because the goal of turbulence is long-time averages of the dynamics rather than temporal accuracy, the iteration to self-consistency across domains, which is typically required by Schwarz schemes, can be avoided, thereby accelerating the computation. Second, we propose a coupling scheme that relies entirely on the addition of source terms, leaving the boundary conditions arbitrary. The practical motivations for such a scheme are discussed, and forms of the source terms that ensure consistency and stability are derived. The schemes are tested on a nonlinear, one-dimensional model problem, and the first scheme is further tested on the Hasegawa–Wakatani model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

On-the-Fly Energy Condensation for Whole-Core Multiphysics Simulations

As the nuclear energy industry considers ways to achieve improved economics in the current fleet of light-water reactors (LWRs), one possible approach is to operate each cycle longer. This causes a greater portion of the fuel to be burned and reduces the frequency of outages, which ultimately reduces the cost to operate the reactor. However, this also leads to higher burnup fuels than have traditionally been allowed in these reactors. There are concerns about integrity of high-burnup (HBu) fuel, especially during accident conditions such as loss-of-coolant accidents (LOCAs), as shown by Capps et al. To investigate these concerns, advanced modeling and simulation capabilities are under development to determine the susceptibility of HBu fuel to fuel fragmentation, relocation, and dispersion (FFRD). Improvements have previously been made to fuel performance capabilities to model these phenomena more accurately; multiphysics simulations have also been conducted to determine the power and burnup histories of the HBu fuel, which are needed as inputs to the fuel performance calculations. Most recently, new statistical approaches have been developed to identify a subset of fuel rods that have greater FFRD susceptibility, reducing the total number of fuel performance simulations required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2004

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Stanojev, B. J.↗
DOE OSTI2015

Cooling, AGN feedback, and star formation in simulated cool-core galaxy clusters

Numerical simulations of active galactic nuclei (AGNs) feedback in cool-core galaxy clusters have successfully avoided classical cooling flows, but often produce too much cold gas. We perform adaptive mesh simulations that include momentum-driven AGN feedback, self-gravity, star formation, and stellar feedback, focusing on the interplay between cooling, AGN heating, and star formation in an isolated cool-core cluster. Cold clumps triggered by AGN jets and turbulence form filamentary structures tens of kpc long. This cold gas feeds both star formation and the supermassive black hole (SMBH), triggering an AGN outburst that increases the entropy of the intracluster medium (ICM) and reduces its cooling rate. Within 1–2 Gyr, star formation completely consumes the cold gas, leading to a brief shutoff of the AGN. The ICM quickly cools and redevelops multiphase gas, followed by another cycle of star formation/AGN outburst. Within 6.5 Gyr, we observe three such cycles. There is good agreement between our simulated cluster and the observations of cool-core clusters. ICM cooling is dynamically balanced by AGN heating, and a cool-core appearance is preserved. The minimum cooling time to free-fall time ratio typically varies between a few and ≳20. The star formation rate (SFR) covers a wide range, from 0 to a few hundred M{sub ⊙} yr{sup −1}, with an average of ∼40 M{sub ⊙} yr{sup −1}. The instantaneous SMBH accretion rate shows large variations on short timescales, but the average value correlates well with the SFR. Simulations without stellar feedback or self-gravity produce qualitatively similar results, but a lower SMBH feedback efficiency (0.1% compared to 1%) results in too many stars.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2017

Improvements to PHISICS/RELAP5-3D{sup C} Capabilities for Simulating HTGRs

Accurate, high-resolution full core simulations of prismatic high temperature reactors for burnup and transient analysis remain a challenge. The cores are geometrically large, but some features are quite small (tri-structural isotropic (TRISO) particles, burnable poison pellets). The treatment of both neutron scattering in graphite and resonance capture are complex and not adequately captured using the methods traditionally used in High Temperature Gas-cooled Reactors (HTGR) and commonly used in Light Water Reactors (LWR). For transient analysis, temperature feedback is a function of the TRISO fuel form, but full core models cannot resolve phenomena at this scale without averaging over space and energy. For burnup calculations, a second level of heterogeneity must be resolved to capture the local effects of fuel and burnable poisons, while accurately propagating their effects through and between blocks. The HTGR Methods and Simulation group has been utilizing the INL-developed PHISICS/RELAP5-3D{sup C} (P/R), code suite for participation in the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Gas Cooled Reactor (MHTGR) Benchmark, funded by the US Department of Energy (DOE) under the Advanced Reactor Technologies (ART) program. The MHTGR-350 benchmark consists of several lattice, steady-state and transient exercises, and the P/R code has been successfully applied to most of the cases defined. In addition to this code-to-code verification activity, it is planned that P/R will be validated using data from several Loss of Forced Cooling (LOFC) tests performed at the High Temperature Test Reactor (HTTR) in a joint effort with Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) project. These verification and validation simulations will identify code and model discrepancies, sensitivities, and establish the current state-of-the-art in HTGR simulation. The Department of Astronautical, Electrical and Energy Engineering (DIAEE), 'Sapienza' University has been involved in the development of P/R since 2014 through various graduate and PhD student projects. In the framework of this collaboration the P/R optimization for application on the OECD MHTGR-350 and HTTR models will be carried out.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3D whole core transport code being developed for the CASL toolset, known as the Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include a subgroup method for resonance self-shielding and a whole-core transport solver with a 2D/1D synthesis method. Depletion capability is supported by coupling SCALE-ORIGEN with MPACT. A new ENDF/B-VII.1 MPACT 51-group library has been developed by using the AMPX/SCALE and CASL-XSTools code packages based on the procedure introduced in the companion paper, “Development of the Multigroup Cross Section Library for the CASL Neutronics Simulator MPACT: Method and Procedure.” The cross section library generation procedure, cross section library, and MPACT transport capability have been verified by performing various benchmark calculations and comparing the benchmark results to the continuous-energy (CE) Monte Carlo results. Furthermore, the code-to-code comparison for the benchmarks shows that the CASL neutronics simulator MPACT and its 51-group library are working reasonably.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Method and procedure

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3D whole core transport code being developed for the CASL toolset known as the Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include a subgroup method for resonance self-shielding and a whole-core transport solver with a 2D/1D synthesis method. The AMPX and SCALE code packages have been significantly improved to include intermediate resonance parameters, as well as new modules to enhance the accuracy of resonance data and scattering matrices for the Bondarenko approach and transport cross sections of hydrogen. In addition, CASL XSTools have been developed to generate the MPACT multigroup cross section library. A simple super-homogenization method for 238U has been developed to resolve reaction rate discrepancy issues caused by angle-dependent total cross sections, poor scattering matrices, and a poor resonance interference model at coarse group structure. Here, due to the limitation of computing capacity that can occur even in high performance computing, a new 51-group structure has been developed for efficient simulation that is applicable for both pressurized water reactor (PWR) and boiling water reactor (BWR) simulations. A new ENDF/B-VII.1 MPACT 51-group library has been developed using the AMPX/SCALE and XSTools code packages, and it has been verified and validated by performing various benchmark calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗