Search NASA⌕ Search

SEARCH · Search NASA

Results for “analysis codes”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Radial Compressor Design and Off-Design for Trans-critical CO2 Operating Conditions

A multi-staged intercooled centrifugal compressor design is performed for an indirect-fired sCO2 cycle and the design point data are used to generate the performance maps using a 2D mean-line analysis code. Fluid and turbomachinery similarity laws are applied to find an appropriate surrogate ideal gas fluid to be used in the design and performance map generation codes. The performance maps are then scaled back for the sCO2 fluid case using the affinity laws. The developed methodology is validated using the experimental data for a compressor operating at similar inflow conditions.

Uysal, Selcuk Can↗
DOE OSTI2022

Assessment of Thermal Stratification in Versatile Test Reactor Transients

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy. The conceptual design of the 300 MWth pool-type sodium-cooled fast reactor (SFR) has been led by the US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. Safety performance analysis for the VTR conceptual design is being performed with the systems thermal-hydraulics (SYSTH) module of the SAS4A/SASSYS-1 liquid-metal reactor safety analysis code system. Since the current model of the VTR employs a simple perfect mixing model for large plena like the hot pool, it is not able to predict temperature variations that may develop during the transient. Prior work simulating the response of SFRs to postulated events like the Protected Station Blackout (PSBO) has shown that the phenomenon of thermal stratification, where stable thermal layers accumulate in the hot pool, may delay the transition to natural circulation and thus impact the predicted transient progression. Thus, an effort has begun to model this transient by integrating a Computational Fluid Dynamics (CFD) model of the hot pool into the SAS4A/SASSYS-1 model of the Primary Heat Transport System during the simulation of the PSBO event. A three-dimensional Volume-Of-Fluid CFD model of the VTR hot pool has been developed for the co-simulation of SAS4A/SASSYS-1 with CFD. In this work, the standalone SAS4A/SASSYS-1 calculation and the standalone CFD calculation based on the SAS4A/SASSYS-1 calculation result were produced on the Idaho National Laboratory High Performance Computing cluster, SAWTOOTH. At this time, only the standalone CFD and SAS4A/SASYS-1 simulation results are provided.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

TRITON11 - Westinghouse 11*11 BWR fuel design

Westinghouse has developed an 11*11 BWR fuel design referred to as the TRITON11 fuel design. Given the current economic pressure on the utilities, the key development objective was to significantly reduce the fuel cycle costs. Careful considerations were also made to meet the varying requirements between the different BWR utilities. In particular, the fuel was optimized for both short and long cycle operation, as well as for the up-rated cores and higher burnups. In addition to these advances, careful considerations were given to the overall reliability of the new design to further reduce the risk of fuel failures. Further improved fuel economy requires improved thermal margins since both SVEA-96 Optima2 and Optima3 are already highly optimized for reactivity performance at the fuel lattice level. It was realized that better thermal margins must come from adding more fuel rods, thereby reducing the linear heat generation rate and surface heat flux. This requires going beyond the 10*10 fuel lattice. An 11*11 fuel design was chosen as optimum, considering benefits versus manufacturing cost and added complexity. The numbers, dimensions and positioning of fuel rods and internal water channels (passing non-boiling water) were analyzed intensively. The resulting TRITON11 design has 109 fuel rods and 3 central water rods for improved moderation. The paper presents the TRITON11 fuel design. The development objectives are laid out and the resulting performance and reliability improvements are described. The development is in the final phase of testing and validation. Examples of recent tests are presented. The advances in BWR core analysis codes in support of TRITON11 fuel are briefly summarized. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Validation of the Modelling of Corrosion Driven Expansion in Grouted Waste-forms - 16125

One of the current preferred solutions for the disposal of Intermediate Level radioactive waste in the UK is grout encapsulation inside steel boxes for long term storage. These boxes are then to be entombed in a Geological Disposal Facility once one is available. However, both prior to and during geological disposal, waste contents can react with the water present in the encapsulant grout and begin to corrode. This corrosion product occupies a larger volume (by several hundred percent) than the original material and the corrosion occurs over the course of decades and centuries, long after the encapsulant grout has set and hardened. The expanding waste then creates stresses in the encapsulant causing both it and the surrounding box to deform and, in extreme circumstances, and after long times, for the structural integrity of the box to fail. This behaviour needs to be quantified before the box can be approved for use as any excessive deformation could lead to a loss of containment or, since the boxes are stacked on top of each other, to a significant accident in the stores. Due to the long timescales involved, physical testing of the phenomenon in full scale is impossible. Instead, a combination of chemical modelling and finite element analysis is used to predict the extent of the corrosion in the waste and its effect of the integrity of the box. There are a number of difficulties in modelling the corrosion-driven expansion and subsequent deformation of the boxes. Although the corrosion occurs over a very long time scale, the event is highly non-linear and dynamic. All of the material models, including the steel and grout are non-linear. The deformation is in the order of 10s of centimetres, well beyond the static elastic range. Finally, the components must be allowed to move relative to each other which involve a lot of part to part contacts. This means that simple static modelling, which would solve quite quickly, cannot be used. Instead a dynamic analysis code, LS-DYNA, has been used to expand the product over time to simulate the corrosion-driven expansion. This software is generally used for modelling car crashes or scenarios where waste containers have been dropped. The analyses therefore include a significant inertial component. This paper will detail the NNL's approach to modelling this long-term, quasi-static phenomenon in the dynamic, impact oriented software code LS-DYNA. It will cover the problems inherent in this approach, the errors that it creates as well as the tests and solutions that NNL worked on in order to overcome them and be confident that the solution could be validated. This work will ultimately be used to determine the maximum inventory which can be stored in each box based on the waste content and the potential extent of expansion as well as any modification that could be made to the box design or the processing streams that may improve the storage solution. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Estimation of Uncertainties of Wall Temperature Accomplished by Model Calibration Technique Using Small to Large Scale Reflooding Test Data

The fidelity of the prediction can be evaluated by performing validation and uncertainty quantification with suitable test data. Nuclear power plant design and operation must accommodate the uncertainties in predicting system performance, where those uncertainties originating due to initial and boundary conditions, numerics, and modeling uncertainties. To improve the prediction of the reactor system performance, these various sources of uncertainties can be refined through model calibration technique. The objective of this work is to refine the model parameters and to estimate the uncertainties on parameters/responses during reflood phase. To accomplish this experiments should be performed, data collected, and the data utilized to complete model calibration for the thermal-hydraulic input parameters. Simulations were performed using Safety and Performance Analysis Code (SPACE) developed by multiple research institutes in South Korea to predict the thermal hydraulic system behaviors of nuclear power plants. Since existing experimental data were obtained from several tests performed at the small scale, scaling considerations must be addressed during the uncertainty quantification process. (authors)

42 ENGINEERING↗
DOE OSTI2016

Conceptual System Design of a Supercritical CO{sub 2} cooled Micro Modular Reactor - 18492

Small modular reactor (SMR) concepts proposed in the past have compact size and a longer life reactor core than the conventional nuclear power plants. However, previous concepts based on the water-cooled technologies with a steam Rankine cycle have shown that the complete modularization is still technically challenging. This study suggests an innovative concept of a reactor directly cooled by supercritical carbon dioxide (S-CO{sub 2}) with S-CO{sub 2} Brayton cycle as a power conversion system, namely KAIST MMR (Micro Modular Reactor). To enhance neutron economy and maximize the fuel volume fraction of the core, a uranium carbide fuel controlled by drum type control rods was used in the fast-spectrum MMR. The core has long life (20 years) without refueling or reshuffling as well as inherent safety features. The reactor can be used as a distributed power source, and replace outdated fossil fuel power plants for small cities. Moreover, we propose a S-CO{sub 2} Brayton cycle as a power conversion system to achieve small, high performance and lightweight module. Thanks to the compact core and power conversion system, the entire system can be contained in a single module. The target of the system is to achieve small and compact system that can be transported via ground transportation of a single module containing the core and the power conversion system altogether. In order to meet this target single module s total weight is minimized in the order of 150 tons. The external size of a module is less than 7 m in length and 4 m in diameter. It produces 12 MWe electricity from 36 MWt reactor core. The S-CO{sub 2} Brayton cycle was optimized, and the cycle components were designed while maximizing the performance and minimizing the weight. A passive decay heat removal system utilizing shell and tube heat exchangers was also developed, and examined its performance with transient analysis code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Evaluation of fission product releases obtained from KueFA safety tests in light of the setup's accuracy - 18692

The Cold Finger Apparatus (KuehlFinger-Apparatur - KueFA) at the European Commission's JRC-Karlsruhe has been designed to assess the effects of Depressurization Loss of Forced Circulation (D-LOFC) accident scenarios on spherical irradiated High Temperature Reactor (HTR) fuel. At approximately 1600 deg. C, the reference maximum temperature for D-LOFC accidents, it has been observed that high-quality German HTR fuel pebbles release only a limited quantity of Cs but a significant amount of Ag. This paper revisits fission product release data obtained from the recently KueFA-tested fuel pebbles HFR-K5/3 and HFR-EU1/3 and presents associated investigations to improve the accuracy of the experimental setup. In light of these findings, implications for previously conducted KueFA heating tests are discussed, considering both experimental data as well as simulated release curves calculated with the Source Term Analysis Code system (STACY). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Computational Modeling, Simulation and Analysis of the RPI Walthousen Reactor Critical Facility (RCF)

The Walthousen Reactor Critical Facility (RCF), located in Schenectady, New York, is currently owned and operated by Rensselaer Polytechnic Institute (RPI). The facility was originally constructed by the American Locomotive Company (ALCO) for the research and development of small, portable reactors. Originally designed for highly-enriched uranium fuel, the prototype reactor first came online in August 26, 1956, and was transferred to RPI ownership in 1963. Several changes to the reactor's design and operation have been made, the most notable of which was the conversion from high-enriched uranium to low-enriched uranium fuel. The RCF is an open-pool critical assembly that operates at a low power with a license limit of less than 100 watts. This low operating power permits a much greater flexibility in configuration and operation as compared to other research reactors. Because of this, experiments may be designed that would be unfeasible for research reactors that operate at much higher powers. The RCF's functional capabilities provide opportunities to perform several reactor experiments and measurements, such as: core criticality experiments, differential and integral rod worth measurements, moderator temperature coefficient of reactivity calculations, critical rod height and critical water height measurements, etc.. These experiments are regularly performed in conjunction with RPI course-work education and Senior Reactor Operator (SRO) training. The unique capabilities and experiments that may be performed in the RCF are particularly useful for validating certain aspects of computational analysis codes. A computational model of the RCF reactor has been previously generated and validated using the Monte Carlo N-Particle (MCNP) code. For the purpose of future experiments, this model is also being adapted for use with the SHARP Toolset, which includes the PROTEUS neutronics code, the Nek5000 fluid dynamics code, and the Diablo structural mechanics code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Characterization of the Pulsed Radiation Environment in FREC-II at the ACRR

Ongoing tests at the Annular Core Research Reactor (ACRR) at Sandia National Laboratories aim to measure temperature and strain in materials exposed to a transient neutron-photon environment. Radiation transport and finite element analysis codes are used to predict the thermal-structural response of test articles exposed in the reactor. Strain and temperature sensors fielded on test articles provide validation data for finite element models, while active and passive radiation dosimetry provides data for MCNP models. The ACRR is able to provide pulse energies of 40 - 300 MJ per pulse with pulse times of 7 - 100 milliseconds. The pulse power is higher than many of the other research reactors due to its unique high heat capacity UO{sub 2}-BeO fuel in the central reactor. Additionally, the Fuel Ringed External Cavity (FREC-II) contains U-ZrH fuel. Future tests at the ACRR plan to field test assemblies in the Fuel Ringed External Cavity (FREC-II) and expose the test articles to the transient anisotropic radiation field experienced in the cavity. The FREC-II is coupled with the ACRR core and is used to test larger specimens, and the radiation environment can be shaped/modified to the needs of the experiment. The FREC-II was designed to provide an irradiation cavity with a strong flux gradient favoring the central cavity side. Neutron fluence also varies by height in the cavity, with a peak level slightly below the fuel rod vertical centerline. Tests were conducted in 2014 to measure the thermal-structural response of test articles exposed in the central cavity. Those tests in the central cavity fielded Nickel activation foils to record the neutron fluence at various locations around the test assembly. Foil data was later used to highlight uncertainties between thermo-structural response models and the measured temperature and strain. Studies performed in 2015 exposed large arrays of activation foils and thermo-luminescent dosimeters (TLD) in the FREC-II cavity in order to map out the neutron and photon spectrum at many locations. This type of passive measurement was shown to be too simple for performing thermo-structural simulations of a test article in response to multiple pulsed-reactor exposures. The Depleted Uranium Strain and Temperature - Enhanced Response (DUSTER-II) experiments will expose test assemblies to reactor pulses in the FREC-II cavity. To prepare for these experiments, the MCNP model will be compared with June 2016 experimental results from the pulses with the FREC-II coupled with the ACRR. These comparisons will provide an accurate model to be implemented in further studies. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Study on Neutronics Calculation Method of SCWR with Annular Fuel Assembly

Supercritical Water-Cooled Reactor (SCWR) is one of Gen-IV reactor systems for its advantages over LWRs, such as high thermal efficiency, simplified reactor system and economics. Annular fuel was proposed by MIT to supplant conventional solid fuel and improve power density in LWRs. Annular fuel is cooled from both inside and outside of the fuel. Thus a lower fuel temperature and cladding surface temperature can be achieved with higher power density. SCWRs will benefit from this advantage because high cladding temperature (usually over 650 degree C) is one challenge of SCWRs. Using annular fuel in SCWRs is very promising. This is also because the inner hole of the annual fuel can provide extra neutron moderation. Supercritical water density is too low to provide enough moderation. Different analysis codes have been developed all over the world to carry out SCWR pre-conceptual core design. Yamaji et al. developed SRAC/ASMBURN/COREBN to study Japan version of SCWR. Monti et al. used ERANOS to analyze HPLWR (High Performance Light Water Reactor). WIMS-D4/CITATION was also used to calculate HPLWR by Tashakor et al.. These codes employ a two-step method used in PWRs calculation. Lattice/Assembly calculation is carried out as the first step. In the second step, few group constant of different fuel assembly types is then used in core calculation. MCNP had also been used to study SCWR by Watta et al.. Due to large computation time cost, it's used mostly to analyze fuel assembly or core performance rather than pre-conceptual design, in which much optimization is needed. In this research, a neutronics calculation code for SCWR is developed using two-step method. This code is verified by comparing with Monte-Carlo code. Fuel assembly homogenization method is studied to improve the code accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Computational Study of Droplet Dynamics in Dispersed Flow Film Boiling Regime in PWR Geometries

Dispersed flow film boiling (DFFB) regime is exhibited during low-flooding rate conditions following the postulated Loss-of-Coolant Accident (LOCA) in nuclear reactor flow channels. LOCA is a type of design-basis accident scenario, which is crucial for reactor design and licensing. Accurate models to predict the heat transfer in DFFB regime is critical for the reactor to maintain safety limits. The nominal operating linear heat rate is also set based on the maximum linear heat rate determined for a LOCA transient. It is understood that the behavior of entrained droplets in DFFB regime greatly affects the heat transfer by evaporating and desuperheating the surrounding steam. Furthermore, interaction with flow structures also aids in increasing the evaporation rate of entrained droplets. In order to quantify this enhancement in heat transfer, the size of droplets downstream of flow structures (Spacer Grids and Mixing Vanes) is of crucial importance. Thermal hydraulic analysis codes like COBRA-TF, CATHARE-3 and adopt a three-field equation model that allows them to model the droplet field separately. However, to ensure the accuracy of these field equations, various correlations and physical models are required. This paper proposes a method for characterizing the mean Sauter Diameter of droplets across spacer-grids and mixing vanes in PWR sub-channels using PHASTA - a high fidelity flow solver with multiphase capabilities using the level-set method. The PHASTA code is a finite element based computational fluid dynamics code designed for solving compressible and incompressible turbulent flows. It has been demonstrated to be massively scalable (up to 786,432 cores), which is crucial for simulating large realistic domains. The code incorporates the level-set method for interface tracking of the gas-liquid interface.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Study on neutronics calculation method of SCWR with annular fuel assembly - 26804

Many Supercritical Water Reactor designs have been proposed all over the world. Annular fuel was employed in SCWR by Xi'an JiaoTong University to achieve lower fuel center temperature and lower cladding surface temperature. Different analysis codes have been developed for these designs. In this article, a neutronics calculation code for SCWR with annular fuel is developed. This code is compared with Monte-Carlo code doing both assembly and core calculation. The results showed that the spectra difference of PWRs and SCWRs is the main reason for assembly calculation errors and conventional assembly homogenization method is not suitable for SCWRs. Thus the group-collapsing and spatial-averaging method were studied to improve the accuracy. The few-group structure which contains 14 energy groups is used. The 'super-assembly' model is employed to consider effect of reflector. The results showed that the accuracy of effective multiplication factor and radial power was improved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

AMPX: A Modern cross section processing system for generating nuclear data libraries - 14619

Nuclear data libraries are essential for nuclear analysis code systems like SCALE. AMPX has been continuously developed since the early 1970's and is used to generate these data libraries for SCALE. Recently, Oak Ridge National Laboratory (ORNL) initiated a major modernization effort to position AMPX for continued development and deployment on modern computing platforms. An important step toward modernization has been to merge AMPX under the SCALE continuous integration (CI) development framework. As a result, AMPX is now developed under the SCALE software quality assurance plan, providing increased confidence in the quality of data libraries developed and deployed by AMPX. This merge allowed us to develop novel methods to generate self-shielding factors that combine SCALE's resonance self-shielding capabilities with the AMPX library-generation capabilities. AMPX processes Evaluated Nuclear Data Format (ENDF/B) files. The ENDF community is working to develop a more modern format that will replace the existing ENDF/B format. Due to strict space and memory requirements when AMPX was first developed, the original AMPX processing and reading of ENDF data were tightly integrated. On newer computer systems, this is no longer necessary. Now there is a separation between reading and processing which allows for simultaneous support of current and new ENDF formats. As a result, AMPX is positioned to adopt any new ENDF format when available. For end users, the modernization effort has resulted in improved processing capabilities for generating nuclear data libraries. The objective of the paper is to document the latest AMPX cross-section processing capabilities and provide the current status of the AMPX modernization effort. (authors)

42 ENGINEERING↗
DOE OSTI2022

Implementation of hybrid finite element method based transport solver in GRIFFIN

A new transport solver option based on the hybrid FEM (HFEM) was implemented in GRIFFIN, the MOOSE-based reactor analysis code, as an effort to support routine core design calculations for advanced reactor applications. The HFEM formulation with P{sub N} (spherical harmonics expansion), akin to the variational nodal method, is effective for solving a spatially homogenized problem with strong transport effect. The residual and Jacobian evaluations of the HFEM weak form were derived and successfully implemented in GRIFFIN, having the diffusion and the PN options available in the new HFEM based transport solver. The performance was tested with the simplified ABTR benchmark problems. The results indicate that the HFEM-based transport solver is a feasible option for solving problems with spatially homogenized and strong streaming by providing superior accuracy with a proper p-refinement. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Energy deposition model in STREAM with photon transport

Recent efforts have been taken to implement a photon transport module for the neutron transport analysis code STREAM, considering the explicit photon energy distribution in the problem instead of using the locally deposition assumption as employed in previous STREAM versions. The current energy release model in STREAM based on the On-the-fly energy released per fission model is modified accordingly to account for the photon energy deposition computed explicitly by this module. The impacts of the modified energy release model with photon transport in STREAM on the power distribution, fuel temperature and other core parameters during depletion with feedback calculations are studied on several problems from the VERA benchmark suit. Overall, the photon transport provides a better power map for depletion and feedback calculations, especially when strong gamma emitters are present. The fuel temperature, in general, is decreased when the photon transport is employed because fission gammas have been transported out of their birthplace. The new energy deposition model in STREAM with photon transport indicates that gamma energy accounts for about 9.7% of the total energy released in a fresh VERA5 2D core and rises to 10.5% at the End of Cycle. Around 2.5% of the total energy released will be deposited in the coolant in this 2D core. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Advances in Metallic Fuel Database Development and Data Qualification

The Fuels Irradiation and Physics Database (FIPD [1]) is a comprehensive repository of data and documents related to Uranium-Zirconium based metallic fuel test pins. This database stores operational conditions of these pins, calculated using a suite of Argonne National Laboratory analysis codes developed during the Integral Fast Reactor (IFR) program. Key calculated data include axial distributions of power, temperature, fluence, burnup, and isotopic densities. Additionally, the FIPD holds post-irradiation examination (PIE) data such as fission gas release, gas chemistry measurements, and axial distributions derived from profilometry, gamma scanning, and neutron radiography. Complementing these data is an extensive archive of documents related to various pins and experiments. These include raw PIE records, design details, safety analyses, and operational reports. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The recent improvements to the database and the data QA status are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiological Safety Analysis Computer (RSAC 7.2) Presentation

This presentation will be for the graduate seminar class for the University of New Mexico. We will discuss the Radiological Safety Analysis Computer (RSAC) version 7.2, a critical tool in radiological consequence analysis within the nuclear safety domain. We will begin with an introduction to radiological consequence analysis and the importance of dose modeling in nuclear safety, referencing key regulatory frameworks such as DOE-STD-3009 and 10 CFR 20. The presentation will provide an overview of RSAC, including its history, primary functions, and the inputs and outputs involved in its calculations. A detailed examination of RSAC's atmospheric dispersion model, based on the Gaussian plume methodology, will be presented, highlighting how meteorological factors influence the results. We will discuss the regulatory alignment of RSAC with DOE and NRC guidelines, emphasizing the assumptions required for compliance and the balance between conservative and realistic inputs. The presentation will also address the limitations and assumptions inherent in RSAC, comparing it with other consequence analysis codes like MACCS, HotSpot, and RASCAL. An example RSAC run will be demonstrated, showcasing a hypothetical accident scenario and interpreting the resulting dose calculations. Finally, we will introduce new ideas to enhance the current RSAC framework. This future work aims to provide multi-industry adaptability and integration with new dispersion models, addressing the evolving needs of DOE and NRC. The presentation will conclude with a discussion on RSAC's ongoing role in nuclear safety and the opportunities for future innovation.

computer program↗
NASA NTRS1995

Development of advanced structural analysis methodologies for predicting widespread fatigue damage in aircraft structures

NASA is developing a 'tool box' that includes a number of advanced structural analysis computer codes which, taken together, represent the comprehensive fracture mechanics capability required to predict the onset of widespread fatigue damage. These structural analysis tools have complementary and specialized capabilities ranging from a finite-element-based stress-analysis code for two- and three-dimensional built-up structures with cracks to a fatigue and fracture analysis code that uses stress-intensity factors and material-property data found in 'look-up' tables or from equations. NASA is conducting critical experiments necessary to verify the predictive capabilities of the codes, and these tests represent a first step in the technology-validation and industry-acceptance processes. NASA has established cooperative programs with aircraft manufacturers to facilitate the comprehensive transfer of this technology by making these advanced structural analysis codes available to industry.

Harris, Charles E.↗