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At least 649 records · Page 36

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NASA NTRS2016

Deductive Evaluation: Formal Code Analysis With Low User Burden

We describe a framework for symbolically evaluating iterative C code using a deductive approach that automatically discovers and proves program properties. Although verification is not performed, the method can infer detailed program behavior. Software engineering work flows could be enhanced by this type of analysis. Floyd-Hoare verification principles are applied to synthesize loop invariants, using a library of iteration-specific deductive knowledge. When needed, theorem proving is interleaved with evaluation and performed on the fly. Evaluation results take the form of inferred expressions and type constraints for values of program variables. An implementation using PVS (Prototype Verification System) is presented along with results for sample C functions.

Di Vito, Ben. L↗

NASA NTRSDate not supplied

Modeling Radiation Sources for Experimental Validation

To build on previously completed experimental work, I have been working to implement a model of our lab’s Americium-Beryllium neutron source and Ludlum detection instruments computationally to provide secondary verification of our experimental results. This has been done using the radiation transport code MCNP6 (Monte Carlo N-Particle). I have developed files to calculate the mass ratio of various composite materials that we have exposed to the radiation source, as well as the components of the detection instruments and converted all of these into geometric coordinates that define the experimental setup. Identifying the proper energy distribution and detector efficiencies as well as the correct interpretation of the detector physics have proven to be the most difficult and time-consuming challenges of the work. It is currently too early for results based on the most updated and accurate models. With this experience, the next step is to develop another source model of a proton beam to simulate exposures on other composite targets. This will be developed based on existing experimental datasets, but not used to validate them. Each of these projects will further define the radiation shielding needs for humans in space radiation environments that astronauts will someday be exposed to beyond Low Earth Orbit.

Juliana Simon↗

NASA NTRSDate not supplied

Proof Mate: An Interactive Proof Helper for PVS (Tool Paper)

This paper presents Proof Mate, an interactive proof helper for the PVS verification system. The helper is integrated in VSCode-PVS, the Visual Studio Code extension for PVS. It extends the capabilities of VSCode-PVS by introducing new functionalities for suggesting proof commands, sketching proof attempts, and repairing broken proofs during interactive proof sessions. This work further aligns VSCode-PVS to the functionalities provided by modern development tools, with the ultimate aim to facilitate the adoption of formal methods in engineering practices and education.

Interactive Theorem Proving↗

NASA NTRSDate not supplied

Proof Mate: An Interactive Proof Helper for PVS

Formal Methods↗

NASA NTRSDate not supplied

Magnetic Suspension Wind Tunnel Reconstruction Using an Extended Kalman Filter Framework

A Kalman filter tool has been created for processing data from NASA Langley’s Magnetic Suspension and Balance System. The filter is formulated to estimate aerodynamic parameters of a model that is levitated magnetically in the test section of the wind tunnel. The Kalman filter tool is a modification of an existing code that has been in use for solving trajectory reconstruction problems and has been validated through previous use supporting many flight projects. Modifications to the code were implemented to add the capability to process data from the magnetic suspension wind tunnel. In particular, the main modifications were to the equations of motion to add models for magnetic and aerodynamic forces and moments. The code has been tested using simulation data to provide a known truth for verification.

Christopher D. Karlgaard↗

DOE OSTI2022

Generating Models of the Flattop Critical Assembly for Benchmark Experiments with Python

Los Alamos National Laboratory has been performing nuclear criticality experiments since 1946 at the Pajarito site, starting the Los Alamos Critical Experiments Facility in 1948. A transition period occurred between 2004 and 2011 as operations moved to the National Criticality Experiments Research Center (NCERC), where criticality experiments are now performed. Criticality experiments are essential for determination and verification of nuclear data used in calculations and modeling—such as radiation transport codes—throughout the industry, enhancing nuclear criticality safety. In addition to nuclear data validation and benchmarking, the remotely operated critical assemblies at NCERC are used for a variety of experiments and training classes supporting criticality safety.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

NASA NTRS2010

Structural Verification and Modeling of a Tension Cone Inflatable Aerodynamic Decelerator

Verification analyses were conducted on membrane structures pertaining to a tension cone inflatable aerodynamic decelerator using the analysis code LS-DYNA. The responses of three structures - a cylinder, torus, and tension shell - were compared against linear theory for various loading cases. Stress distribution, buckling behavior, and wrinkling behavior were investigated. In general, agreement between theory and LS-DYNA was very good for all cases investigated. These verification cases exposed the important effects of using a linear elastic liner in membrane structures under compression. Finally, a tension cone wind tunnel test article is modeled in LS-DYNA for which preliminary results are presented. Unlike data from supersonic wind tunnel testing, the segmented tension shell and torus experienced oscillatory behavior when subjected to a steady aerodynamic pressure distribution. This work is presented as a work in progress towards development of a fluid-structures interaction mechanism to investigate aeroelastic behavior of inflatable aerodynamic decelerators.

Tanner, Christopher L.↗

DOE OSTI2022

CTF Validation and Verification: Version 4.3

Coolant-Boiling in Rod Arrays- Two Fluids (COBRA-TF) is a thermal/hydraulic (T/H) simulation code designed for light water reactor (LWR) analysis. It uses a two-fluid, three-field (i.e., fluid film, fluid drops, and vapor) modeling approach. Both subchannel and 3D Cartesian forms of nine conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and has been used and modified by several institutions over the last several decades. COBRA-TF is also used at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Modeling Group (RDFMG) and has been improved, updated, and subsequently became the CTF. One part of the improvement process includes validating the methods in CTF. This document seeks to provide a certain level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model—rod bundle geometries with operating conditions that are representative of prototypical pressurized water reactor (PWR)s and boiling water reactor (BWR)s in both normal and accident conditions. This is done by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

CTF Validation and Verification: Version 4.4

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Study of the heat pipe vapor flow using the MOOSE application Sockeye and the CFD code Nek5000

Heat pipes are efficient heat transfer devices used in various applications, including nuclear microreactors. Heat pipe-cooled microreactors offer advantages in size and cost, which can significantly accelerate their deployment and adoption. Developing accurate models of heat pipes is crucial for a heat pipe-cooled microreactor design and operation. In this work a verification and validation of the heat pipe application Sockeye is presented. Results are compared against RANS simulations, performed with the code Nek5000, and the SPHERE experiment, conducted at INL.

42 ENGINEERING↗

NASA NTRS1983

The effect of incidence angle on the overall three-dimensional aerodynamic performance of a classical annular airfoil cascade

To be of quantitative value to the designer and analyst, it is necessary to experimentally verify the flow modeling and the numerics inherent in calculation codes being developed to predict the three dimensional flow through turbomachine blade rows. This experimental verification requires that predicted flow fields be correlated with three dimensional data obtained in experiments which model the fundamental phenomena existing in the flow passages of modern turbomachines. The Purdue Annular Cascade Facility was designed specifically to provide these required three dimensional data. The overall three dimensional aerodynamic performance of an instrumented classical airfoil cascade was determined over a range of incidence angle values. This was accomplished utilizing a fully automated exit flow data acquisition and analysis system. The mean wake data, acquired at two downstream axial locations, were analyzed to determine the effect of incidence angle, the three dimensionality of the cascade exit flow field, and the similarity of the wake profiles. The hub, mean, and tip chordwise airfoil surface static pressure distributions determined at each incidence angle are correlated with predictions from the MERIDL and TSONIC computer codes.

Bergsten, D. E.↗

DOE OSTI2025

Improvement and Verification of Online Cross Section Generation Capability of Griffin for TRISO-fueled Reactors

Griffin, a MOOSE-based reactor multiphysics code jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE Office of Nuclear Energy’s NEAMS program, has pursued the development of an online multigroup cross section generation capability for a few years to enable high-fidelity, problem-dependent neutronics analyses of advanced thermal reactors. Recent advancements in Griffin’s online multigroup cross section generation capability have significantly improved the accuracy, robustness, and efficiency of self-shielding calculations for both prismatic and pebble-bed TRISO-fueled reactor applications. Key developments include a unified fuel self-shielding method applicable to both TRISO and annular compact/spherical shell fuel zone geometries; an advanced Dancoff Category-based Equivalence Theory using a bell function for non-fuel resonance treatment, achieving more than an order-of-magnitude speedup compared to the Tone method; an on-the-fly multigroup equivalence approach to mitigate group condensation errors; and a streaming correction method for pebble-bed homogenization. A proof-of-concept demonstration of on-the-fly group condensation with consistent P0 transport correction was also achieved. The method reproduced direct fine-group solutions with excellent accuracy (eigenvalue errors within 10 pcm and pin-power differences within 0.5%), but due to performance limitations of the current fixed-source solver, improvements to solver efficiency will be addressed in future work. Verification tests were performed on graphite-moderated TRISO-fueled two-dimensional core benchmark problems representing gas-cooled microreactors, heat pipe-cooled microreactors, gas-cooled pebble-bed reactors, and fluoride salt-cooled high-temperature reactors. Across all cases, Griffin showed excellent agreement with Serpent2 continuous energy Monte Carlo solutions: eigenvalue errors within 200 pcm, pin-power root-mean-square errors within 2%, and control rod and drum worth errors less than 2%. It should be noted that, for the benchmark problem, cross section generation contributed less than 3% of the total simulation times. These results demonstrate that Griffin’s online cross section generation capability delivers accurate and efficient reactor physics solutions across a wide spectrum of TRISO-fueled advanced reactor designs. With further improvements to the fine-group fixed-source solver and planned extensions to depletion, transients, and coupled neutron–gamma transport, Griffin will be well-positioned to become a powerful and comprehensive tool for advanced reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1988

Computer codes for thermal analysis of a solid rocket motor nozzle

A number of computer codes are available for performing thermal analysis of solid rocket motor nozzles. Aerotherm Chemical Equilibrium (ACE) computer program can be used to perform one-dimensional gas expansion to determine the state of the gas at each location of a nozzle. The ACE outputs can be used as input to a computer program called Momentum/Energy Integral Technique (MEIT) for predicting boundary layer development development, shear, and heating on the surface of the nozzle. The output from MEIT can be used as input to another computer program called Aerotherm Charring Material Thermal Response and Ablation Program (CMA). This program is used to calculate oblation or decomposition response of the nozzle material. A code called Failure Analysis Nonlinear Thermal and Structural Integrated Code (FANTASTIC) is also likely to be used for performing thermal analysis of solid rocket motor nozzles after the program is duly verified. A part of the verification work on FANTASTIC was done by using one and two dimension heat transfer examples with known answers. An attempt was made to prepare input for performing thermal analysis of the CCT nozzle using the FANTASTIC computer code. The CCT nozzle problem will first be solved by using ACE, MEIT, and CMA. The same problem will then be solved using FANTASTIC. These results will then be compared for verification of FANTASTIC.

Chauhan, Rajinder Singh↗

DOE OSTI2024

Extended-MHD simulations of disruption mitigation via massive gas injection in SPARC

Recent developments to the M3D-C1 code enable higher fidelity modeling of disruptions, and can be applied in the design verification of reactor-scale tokamaks. Among these new capabilities is a method to mesh conducting vessel structures such as coils and passive plates, packing of the toroidal mesh around gas injectors, as well as anisotropic resistivity inside the vessel structures. We present extended-magnetohydrodynamic (MHD) simulations of disruption mitigation via massive gas injection (MGI) in SPARC. The goal of this study is to inform the disruption mitigation layout of SPARC and aid in the design of an effective gas injector configuration. Fully three-dimensional simulations with M3D-C1 are carried out for various injector configurations with the primary goal of determining the effect of different MGI parameters on heat loads and vessel forces. The simulations include a model for impurity ionization, recombination, advection and radiation, as well as spatially resolved conducting structures around the plasma. A localized mixture of deuterium and neon with a small toroidal and poloidal width is injected in up to six locations. We demonstrate that M3D-C1 can model a rapid shutdown via MGI using narrow and more realistic gas plumes than in previous simulations. As a result of the q = 1 surface in the SPARC baseline case a sawtooth is observed early in the simulations. Despite the sawtooth and the onset of edge MHD instabilities, the impurity distribution remains localized around the injector locations, but enables a radiative shutdown of the plasma. We find that using the maximum of six gas injectors results in a lower peaking factor and leads to a more even distribution of radiation toroidally than using two injectors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

(U) SENSMG: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate Ratios, Leakage, k eff , α , and Subcritical Multiplication Using PARTISN

SENSMG is a tool for computing first-order sensitivities of neutron reaction rates, reaction-rate ratios, leakage, k eff , α, and subcritical multiplication using the PARTISN multigroup discrete-ordinates code. SENSMG computes sensitivities to all of the transport cross sections and data (total, fission, a nu, chi, and all scattering moments), two edit cross sections (absorption and capture), and the density for every nuclide and energy group. It also computes sensitivities to the mass density for every material and derivatives with respect to all interface locations and outer boundaries. It computes sensitivities to user specified reactions whose cross sections are available in a user-supplied NJOY output file. The tool can be used for one-dimensional spherical and slab (r) and two-dimensional cylindrical (r-z) geometries. The tool can be used for fixed-source and eigenvalue problems. For most responses, the tool implements Generalized Perturbation Theory (GPT) as discussed by Williams and Stacey. The tool is thus limited to computing sensitivities only for GPT-allowable responses. For subcritical multiplication, the tool implements sensitivities derived by O’Brien and Clark. SENSMG has a similar role as the old SWANLAKE (Ref. 8), FORSS (Ref. 9), and SENSIT (Ref. 10) codes. It has capabilities similar to those of SUSD3D (Refs. 11 and 12), which also uses PARTISN. Section II of this report describes the theory behind adjoint-based sensitivities, gives the equations that SENSMG solves, and defines the sensitivities that are output. Section III describes the user interface, including the input file and command line options. Section IV describes the output. Section V gives some notes about the coding that may be of interest. Section VI presents some sample problems and discusses verification, which is ongoing. Section VII lists needs and ideas for future work. Appendix A lists most of the input files whose results are presented in Sec. VI. Appendix B provides some useful details on one of the cross-section libraries that SENSMG supports.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2026

Continuous-Energy Verification of MCNP Calculations Using One-Group Spherical and Slab Criticality Benchmarks

This work presents a continuous-energy Monte Carlo verification study of one group spherical and slab criticality benchmarks using the MCNP ®1 code. Classical tabulations and newly generated benchmark solutions obtained by direct numerical evaluation by the author are considered. The benchmarks span weakly to strongly multiplying regimes and provide analytically defined critical radii as functions of a single parameter, c .

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1972

Design and flight experience with a digital fly-by-wire control system using Apollo guidance system hardware on an F-8 aircraft.

This paper discusses the design and initial flight tests of the first digital fly-by-wire system to be flown in an aircraft. The system, which used components from the Apollo guidance system, was installed in an F-8 aircraft. A lunar module guidance computer is the central element in the three-axis, single-channel, multimode, digital, primary control system. An electrohydraulic triplex system providing unaugmented control of the F-8 aircraft is the only backup to the digital system. Emphasis is placed on the digital system in its role as a control augmentor, a logic processor, and a failure detector. A sampled-data design synthesis example is included to demonstrate the role of various analytical and simulation methods. The use of a digital system to implement conventional control laws was shown to be practical for flight. Logic functions coded as an integral part of the control laws were found to be advantageous. Verification of software required an extensive effort, but confidence in the software was achieved. Initial flight results showed highly successful system operation, although quantization of pilot's stick and trim were areas of minor concern from the piloting standpoint.

Deets, D. A.↗

NASA NTRS1978

Computation of the viscous supersonic flow over symmetrical and asymmetrical external axial corners

The primary objective of the reported investigation is the computational verification of the experimental results obtained by Salas and Daywitt (1978). Two existing computer codes were used to compute the supersonic flow field surrounding the external axial corner. For the inviscid and turbulent flow results, the unsteady, three-dimensional implicit code of Pulliam and Steger (1978) was used. For the laminar flow results, the unsteady two-dimensional explicit procedure of Vigneron et al. (1977) was employed. Inviscid solutions for a symmetric configuration with a rounded corner resulted in either single or triple surface crossflow stagnation point flows, depending on the corner radius. Numerical results obtained for the same symmetric configuration tested experimentally show the crossflow in the vicinity of the corner to be away from the corner and thus in agreement with the experimental oil flow results.

Kutler, P.↗