Results for “Code assessment” — Page 19

NASA NTRS2010

Aeromechanics and Aeroacoustics Predictions of the Boeing-SMART Rotor Using Coupled-CFD/CSD Analyses

This paper will highlight helicopter aeromechanics and aeroacoustics prediction capabilities developed by Georgia Institute of Technology, the Pennsylvania State University, and Northern Arizona University under the Helicopter Quieting Program (HQP) sponsored by the Tactical Technology Office of the Defense Advanced Research Projects Agency (DARPA). First initiated in 2004, the goal of the HQP was to develop high fidelity, state-of-the-art computational tools for designing advanced helicopter rotors with reduced acoustic perceptibility and enhanced performance. A critical step towards achieving this objective is the development of rotorcraft prediction codes capable of assessing a wide range of helicopter configurations and operations for future rotorcraft designs. This includes novel next-generation rotor systems that incorporate innovative passive and/or active elements to meet future challenging military performance and survivability goals.

Bain, Jeremy↗

DOE OSTI2019

Status Report on Separate Effects Testing of Thermal Conductivity of U-Zr Alloy

Phases by definition in an alloy system have distinct physical and chemical properties, including thermal conductivity. The goal of this multi-year study is to show that the different phases present in the U-Zr system indeed do have different thermal conductivities. This study will provide meso-scale (10-100 µm) experimental observations using the Thermal Conductivity Microscope (TCM) of phase dependent thermal conductivity. The results of this study may indicate the need to reconsider the application of homogeneous medium assumption for metallic nuclear fuel that is currently used in fuel performance modeling, and could provide input for improving such model using a heterogeneous phase dependent assessment of thermal properties. The local thermal conductivity values from the TCM will be compared to the bulk measurements to validate this work. Twelve alloy samples with varying Zr compositions and heat treatments were fabricated for the initial phase of this study to characterize the individual phases present in the U-Zr system. The gamma body-centered cubic (BCC) phase was successfully identified in the appropriate size for TCM measurement. Next steps in this work will include identification of remaining phases in the U-Zr alloy system. Thermal conductivity will be measured in those phase areas of the materials and used to assess fuel performance codes for improvements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Evaluation of the adequacy of using few-group lattice-homogenized properties for the diffusion analysis of the super critical water reactor

Two issues may affect the accuracy of computed core reactivities and flux/power distributions for the Super Critical Water Reactor (SCWR) core with traditional core-analysis code RFSP: one is the two-energy-group neutron-diffusion theory; the other is the generation of lattice-homogenized properties with the lattice code based on the single-lattice-cell model without considering the effects of the environment. These two issues are not SCWR specific; however their effect may be more significant for SCWR. It has been illustrated that the lattice-homogenized properties calculated with the single-lattice-cell model is not sufficiently accurate for heterogeneous core configurations such as ACR-1000 checkerboard-voiding and core-reflector interface when adjacent channels experience significant spectrum interaction. To evaluate the adequacy of using two-group neutron-diffusion theory with single-lattice-based lattice properties for the analysis of the SCWR core, a 2-D SCWR benchmark problem was setup with the reference solution provided by the continuous-energy Monte-Carlo code SERPENT. The assessment shows that the traditional two-group neutron-diffusion theory with the single-lattice-cell- based lattice properties is not sufficient to capture either the spectral change or the environment effect for the SCWR core. The solution of the eight-group neutron-diffusion equation with the multicell- based lattice properties is considered appropriate for the analysis of the d SCWR core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Analysis of ELAP and mitigation strategies in Maanshan PWR with PCTRAN

On March 11, 2011, the Fukushima Daiichi Nuclear Power Plant, owned by Tokyo Electric Power Company, encountered a beyond design basis earthquake/tsunami and the flood crippled the on-site emergency power and electric transmission systems. The ELAP (Extended Loss of Alternating Current Power) and LUHS (Loss of Ultimate Heat Sink) events severely damaged the core cooling systems and the safety functions of the containment. As a result, three reactor cores were damaged. Also, the cooling capability of nuclear spent fuel pools was lost due to loss of AC and DC power. Existing Station Blackout (SBO) analyses have demonstrated that even with leakage from Reactor Coolant Pump (RCP) seals, the core could maintain coolable only when heat transfer exists between primary loop and secondary loop. Depend on the availability of DC power in different plant designs, the SBO coping times were shown to be 4 to 16 hours. According to INPO event report, it is necessary to analyze the consequences if current coping times are extended with increased core water inventory loss from pump seals. In an ELAP scenario, all active system including Emergency Core Cooling System (ECCS) and motor driven AFW were out of power supply, results in degradation of core cooling capability. Moreover, the fuel exposed risk would increase due to continuously Reactor Coolant System (RCS) inventory loss through pump seals. Therefore, mitigation strategy must be applied prior to core uncovering to prevent its damage and assure plant safety. This research utilized a fast running NPP safety simulation code, PCTRAN, to assess reactor performances in the situations of ELAP of Maanshan PWR. Also, the sensitivity study of injection flow rates and effectiveness of mitigation strategy were demonstrated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Validation of Scale 6.2 criticality calculations using KENO V.A and KENO-VI - 14603

SCALE 6.2 provides enhancements for criticality calculations relative to SCALE 6.1, especially with updated nuclear data libraries, resonance self-shielding options for multigroup (MG) calculations, and problem-dependent temperature treatments for continuous-energy (CE) calculations. Benchmark experiments from the International Criticality Safety Benchmark Evaluation Project that are available in the Oak Ridge National Laboratory Verified, Archived, Library of Inputs and Data (VALID) were applied in this validation effort to examine the impact of the SCALE 6.2 enhancements. Over 300 cases are drawn from the VALID library for testing KENO V.a, and a more limited set of approximately 50 configurations is used for validation of KENO-VI. A total of 401 unique critical experiments are used in this validation effort. In some cases, the KENO V.a models were converted to KENO-VI models so that the performance of both codes could be assessed for the same systems. Validation results for the SCALE 6.2 ENDF/B-VII.1 nuclear data libraries are presented, including MG neutron data in 56- and 252-group structures as well as CE data. The performance of all three multipurpose neutron libraries in calculating k{sub eff} is examined with critical experiment models covering a wide range of fuels, moderators, reflectors, absorbers, and neutron spectra. The SCALE 6.2 ENDF/B-VII.1 coupled neutron-photon library, primarily developed for shielding applications with 200 neutron groups and 47 gamma groups, provides additional energy resolution for fast systems relative to the MG neutron libraries which were developed to produce low biases for thermal neutron systems. The applicability of this library to criticality calculations for fast neutron spectrum systems is not investigated here but will be addressed in the final SCALE 6.2 validation report. The results of these calculations demonstrate that SCALE 6.2 generally provides lower biases than SCALE 6.1 and that the two versions of KENO provide equivalent results on the same suite of problems. SCALE 6.2 also includes an updated neutron cross-section covariance library based on the ENDF/B-VII.1 covariance data. The effect of these updated covariances on the uncertainty in k{sub eff} due to cross section uncertainties is also presented for some cases. A more thorough investigation of the updated covariance data is presented in a companion paper. (authors)

42 ENGINEERING↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Fuel Performance Simulation of High Burnup Fuels in Planned Integral Design Basis Accident Experiments

High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

CIRCEE v1 - a versatile tool for criticality accident dose assessment - 14527

CIRCEE is a calculation code developed by the French Atomic Energy Commission (CEA), allowing fast assessment of operational dose over complex geometries. It is used to assess dosimetric consequences of a criticality accident. More specifically, it can be used both during the preliminary studies phase (isodoses curves for safety report preparation or emergency preparedness, CAAS implantation and validation...) and during crisis management (effective dose estimation, emergency response). An early development stage of CIRCEE has been presented in a talk at the 2009 NCSD edition [1]. Since then, an extensive code rewriting increased its reliability and maintainability while keeping its performance at an excellent level. Now in a stable production version, and following the grant of several international patents, it should be available under commercial license terms in the near future. (author)

42 ENGINEERING↗

DOE OSTI2015

RELAP5-3D Developmental Assessment: Comparison of Versions 4.3.4i and 4.2.1i

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.3.4i and 4.2.1i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1996

Euler Technology Assessment for Preliminary Aircraft Design: Compressibility Predictions by Employing the Cartesian Unstructured Grid SPLITFLOW Code

The objective of the second phase of the Euler Technology Assessment program was to evaluate the ability of Euler computational fluid dynamics codes to predict compressible flow effects over a generic fighter wind tunnel model. This portion of the study was conducted by Lockheed Martin Tactical Aircraft Systems, using an in-house Cartesian-grid code called SPLITFLOW. The Cartesian grid technique offers several advantages, including ease of volume grid generation and reduced number of cells compared to other grid schemes. SPLITFLOW also includes grid adaption of the volume grid during the solution to resolve high-gradient regions. The SPLITFLOW code predictions of configuration forces and moments are shown to be adequate for preliminary design, including predictions of sideslip effects and the effects of geometry variations at low and high angles-of-attack. The transonic pressure prediction capabilities of SPLITFLOW are shown to be improved over subsonic comparisons. The time required to generate the results from initial surface data is on the order of several hours, including grid generation, which is compatible with the needs of the design environment.

Finley, Dennis B.↗

DOE OSTI2025

Quasi-optical beam tracing module development for millimeter-wave high-wavenumber collective scattering on the NSTX-U and EAST tokamaks

A Python3-based beam tracing code utilizing Quasi-Optics has been developed to track both incident and receiving beams in high-k collective millimeter wave scattering systems within magnetic fusion plasmas. In contrast to existing ray tracing codes that solely consider refraction, this beam tracing code incorporates diffraction phenomena, providing a more comprehensive calculation. Here, this enhanced capability allows for a more accurate calculation of the scattering volume and spatial resolution in high-k collective scattering systems, crucial for evaluating system performance and facilitating data analysis. Unlike Geometrical Optics, Quasi-Optics employs the complex eikonal method, representing a Gaussian beam as a collection of coupled rays to accurately preserve diffraction characteristics. The developed code is intended for application in NSTX-Upgrade and EAST high-k beam tracing analyses, targeting frequencies of 693 GHz and 270 GHz, respectively. The high-k system's primary objective is the observation of electron-scale instabilities. Employing a symplectic integrator, the code ensures numerical accuracy, assessed through the conservation of the Hamiltonian. With its precision and efficiency, the code facilitates rapid inter-shot analyses.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

NASA NTRS2008

Assessment of Microphysical Models in the National Combustion Code (NCC) for Aircraft Particulate Emissions: Particle Loss in Sampling Lines

This paper at first describes the fluid network approach recently implemented into the National Combustion Code (NCC) for the simulation of transport of aerosols (volatile particles and soot) in the particulate sampling systems. This network-based approach complements the other two approaches already in the NCC, namely, the lower-order temporal approach and the CFD-based approach. The accuracy and the computational costs of these three approaches are then investigated in terms of their application to the prediction of particle losses through sample transmission and distribution lines. Their predictive capabilities are assessed by comparing the computed results with the experimental data. The present work will help establish standard methodologies for measuring the size and concentration of particles in high-temperature, high-velocity jet engine exhaust. Furthermore, the present work also represents the first step of a long term effort of validating physics-based tools for the prediction of aircraft particulate emissions.

Wey, Thomas↗

NASA NTRS2010

Assessment of Static Delamination Propagation Capabilities in Commercial Finite Element Codes Using Benchmark Analysis

With capabilities for simulating delamination growth in composite materials becoming available, the need for benchmarking and assessing these capabilities is critical. In this study, benchmark analyses were performed to assess the delamination propagation simulation capabilities of the VCCT implementations in Marc TM and MD NastranTM. Benchmark delamination growth results for Double Cantilever Beam, Single Leg Bending and End Notched Flexure specimens were generated using a numerical approach. This numerical approach was developed previously, and involves comparing results from a series of analyses at different delamination lengths to a single analysis with automatic crack propagation. Specimens were analyzed with three-dimensional and two-dimensional models, and compared with previous analyses using Abaqus . The results demonstrated that the VCCT implementation in Marc TM and MD Nastran(TradeMark) was capable of accurately replicating the benchmark delamination growth results and that the use of the numerical benchmarks offers advantages over benchmarking using experimental and analytical results.

Orifici, Adrian C.↗

DOE OSTI2013

RELAP5-3D Developmental Assessment: Comparison of Versions 4.1.3is and 4.0.3is

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.1.3is and 4.0.3is. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

RELAP5-3D Developmental Assessment: Comparison of Versions 4.2.1i and 4.1.3i

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code using versions 4.2.1i and 4.1.3i. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions changed between these two code versions and can be used to identify cases in which the assessment judgment may need to be changed in Volume III of the code manual. Changes to the assessment judgments made after reviewing all of the assessment cases are also provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2010

Comparison of CFD Predictions with Shuttle Global Flight Thermal Imagery and Discrete Surface Measurements

Surface temperature measurements from the STS-119 boundary-layer transition experiment on the space shuttle orbiter Discovery provide a rare opportunity to assess turbulent CFD models at hypersonic flight conditions. This flight data was acquired by on-board thermocouples and by infrared images taken off-board by the Hypersonic Thermodynamic Infrared Measurements (HYTHIRM) team, and is suitable for hypersonic CFD turbulence assessment between Mach 6 and 14. The primary assessment is for the Baldwin-Lomax and Cebeci-Smith algebraic turbulence models in the DPLR and LAURA CFD codes, respectively. A secondary assessment is made of the Shear-Stress Transport (SST) two-equation turbulence model in the DPLR code. Based upon surface temperature comparisons at eleven thermocouple locations, the algebraic-model turbulent CFD results average 4% lower than the measurements for Mach numbers less than 11. For Mach numbers greater than 11, the algebraic-model turbulent CFD results average 5% higher than the three available thermocouple measurements. Surface temperature predictions from the two SST cases were consistently 3 4% higher than the algebraic-model results. The thermocouple temperatures exhibit a change in trend with Mach number at about Mach 11; this trend is not reflected in the CFD results. Because the temperature trends from the turbulent CFD simulations and the flight data diverge above Mach 11, extrapolation of the turbulent CFD accuracy to higher Mach numbers is not recommended.

Wood, William A.↗

DOE OSTI2016

Assessment of a Hybrid Continuous/Discontinuous Galerkin Finite Element Code for Geothermal Reservoir Simulations

FALCON (“Fracturing And Liquid CONvection”) is a hybrid continuous / discontinuous Galerkin finite element geothermal reservoir simulation code based on the MOOSE (“Multiphysics Object-Oriented Simulation Environment”) framework being developed and used for multiphysics applications. In the present work, a suite of verification and validation (“V&V”) test problems for FALCON was defined to meet the design requirements, and solved to the interests of enhanced geothermal system (“EGS”) design. Furthermore, the intent for this test problem suite is to provide baseline comparison data that demonstrates the performance of the FALCON solution methods. The simulation problems vary in complexity from singly mechanical or thermo process, to coupled thermo-hydro-mechanical processes in geological porous media. Numerical results obtained by FALCON agreed well with either the available analytical solution or experimental data, indicating the verified and validated implementation of these capabilities in FALCON. Some form of solution verification has been attempted to identify sensitivities in the solution methods, where possible, and suggest best practices when using the FALCON code.

15 GEOTHERMAL ENERGY↗