Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Data Testing”

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 1,369 records · Page 76
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

What can football tell us about nuclear testing?

When football fans in a stadium cheer loudly enough, it can create a seismic signal. But how do we know cheering is the cause and not something else? Scientists at Los Alamos National Laboratory study how to differentiate between “nuisance alarms” – like “fanquakes,” earthquakes, or mine explosions – and an underground nuclear test. This information helps improve tools for international nuclear test treaty monitoring – and makes the world safer.

58 GEOSCIENCES↗
DOE OSTI2016

Integration Of PanDA Workload Management System With Supercomputers for ATLAS and Data Intensive Science

The.LHC, operating at CERN, is leading Big Data driven scientific explorations. Experiments at the LHC explore the fundamental nature of matter and the basic forces that shape our universe. ATLAS, one of the largest collaborations ever assembled in the sciences, is at the forefront of research at the LHC. To address an unprecedented multi-petabyte data processing challenge, the ATLAS experiment is relying on a heterogeneous distributed computational infrastructure. The ATLAS experiment uses PanDA (Production and Data Analysis) Workload Management System for managing the workflow for all data processing on over 150 data centers. Through PanDA, ATLAS physicists see a single computing facility that enables rapid scientific breakthroughs for the experiment, even though the data centers are physically scattered all over the world. While PanDA currently uses more than 250,000 cores with a peak performance of 0.3 petaFLOPS, LHC data taking runs require more resources than grid can possibly provide. To alleviate these challenges, LHC experiments are engaged in an ambitious program to expand the current computing model to include additional resources such as the opportunistic use of supercomputers. We will describe a project aimed at integration of PanDA WMS with supercomputers in United States, in particular with Titan supercomputer at Oak Ridge Leadership Computing Facility. Current approach utilizes modified PanDA pilot framework for job submission to the supercomputers batch queues and local data management, with light-weight MPI wrappers to run single threaded workloads in parallel on LCFs multi-core worker nodes. This implementation was tested with a variety of Monte-Carlo workloads on several supercomputing platforms for ALICE and ATLAS experiments and it is in full pro duction for the ATLAS since September 2015. We will present our current accomplishments with running PanDA at supercomputers and demonstrate our ability to use PanDA as a portal independent of the computing facilities infrastructure for High Energy and Nuclear Physics as well as other data-intensive science applications, such as bioinformatics and astro-particle physics.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Preliminary description of a new creep-fatigue design method that reduces over conservatism and simplifies the high temperature design process

The report provides the initial description of a new creep-fatigue design method for structural components in high temperature nuclear service. The new method is based on an integrated elastic-perfectly plastic (EPP) analysis and Simplified Model Test (SMT) approach that reduces over conservatism, improves the treatment of elastic follow up, and simplifies the design procedure, when compared with the current creep-fatigue design methods in ASME Boiler and Pressure Vessel Code. Developing the design charts for the EPP-SMT design method requires extrapolating SMT test data as a function of hold time and follow up factor. The report develops the preliminary design charts for Alloy 617 at temperatures between 800°C and 950°C by combining two extrapolation approaches developed in a previous work. The report also presents a comparative analysis between the EPP-SMT design method and the current ASME creep-fatigue design methods by evaluating design life of two sample geometries under different loading conditions. Results from the comparative analysis verify the EPP-SMT design charts but suggest the requirement of additional test data in the low strain range regime for improving the extrapolation procedure that will further reduce the over conservatism in the creep-fatigue damage evaluation. The report also concludes that the EPP-SMT design procedure can account for effect of primary load on creep-fatigue life by using a fixed, bounding value of follow up in constructing the design charts. The conclusions to this report describe the future work required to complete this new design method so it can be codified through a nuclear Code Case.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Nuclear level densities of 64 , 66 Zn from neutron evaporation

Double differential cross sections of neutrons from d+ 63,65 Cu reactions have been measured at deuteron energies of 6 and 7.5 MeV. The cross sections measured at backward angles have been compared to theoretical calculations in the framework of the statistical Hauser-Feshbach model. Three different level density models were tested: the Fermi-gas model, the Gilbert-Cameron model, and the microscopic approach through the Hartree-Fock-Bogoliubov method (HFBM). The calculations using the Gilbert-Cameron model are in best agreement with our experimental data. Level densities of the residual nuclei 64 Zn and 66 Zn have been obtained from statistical neutron evaporation spectra. In conclusion, the angle-integrated cross sections have been analyzed with the exciton model of nuclear reaction.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Corrective Action Decision Document/Corrective Action Plan for Corrective Action Unit 97: Yucca Flat/Climax Mine Nevada National Security Site, Nevada, Revision 1

This corrective action decision document (CADD)/corrective action plan (CAP) has been prepared for Corrective Action Unit (CAU) 97, Yucca Flat/Climax Mine, Nevada National Security Site (NNSS), Nevada. The Yucca Flat/Climax Mine CAU is located in the northeastern portion of the NNSS and comprises 720 corrective action sites. A total of 747 underground nuclear detonations took place within this CAU between 1957 and 1992 and resulted in the release of radionuclides (RNs) in the subsurface in the vicinity of the test cavities. The CADD portion describes the Yucca Flat/Climax Mine CAU data-collection and modeling activities completed during the corrective action investigation (CAI) stage, presents the corrective action objectives, and describes the actions recommended to meet the objectives. The CAP portion describes the corrective action implementation plan. The CAP presents CAU regulatory boundary objectives and initial use-restriction boundaries identified and negotiated by DOE and the Nevada Division of Environmental Protection (NDEP). The CAP also presents the model evaluation process designed to build confidence that the groundwater flow and contaminant transport modeling results can be used for the regulatory decisions required for CAU closure. The UGTA strategy assumes that active remediation of subsurface RN contamination is not feasible with current technology. As a result, the corrective action is based on a combination of characterization and modeling studies, monitoring, and institutional controls. The strategy is implemented through a four-stage approach that comprises the following: (1) corrective action investigation plan (CAIP), (2) CAI, (3) CADD/CAP, and (4) closure report (CR) stages.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Product Consistency Test Results for the LAW ALG Glasses

This report provides the results of the Product Consistency Test leachates from the Low-Activity Waste Algorithm glasses, a series of simulated nuclear waste glasses designed and fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister centerline cooled versions of the glasses. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1991

Rover nuclear rocket engine program: Overview of rover engine tests

The results of nuclear rocket development activities from the inception of the ROVER program in 1955 through the termination of activities on January 5, 1973 are summarized. This report discusses the nuclear reactor test configurations (non cold flow) along with the nuclear furnace demonstrated during this time frame. Included in the report are brief descriptions of the propulsion systems, test objectives, accomplishments, technical issues, and relevant test results for the various reactor tests. Additionally, this document is specifically aimed at reporting performance data and their relationship to fuel element development with little or no emphasis on other (important) items.

Finseth, J. L.↗
DOE OSTI2016

Microstructural Characterization of In-reactor Corrosion Tested Hastelloy N{sup R} and 316 Stainless Steel

The success of the Molten Salt Reactor Experiment (MSRE) at the Oak Ridge National Laboratory (ORNL) in the period 1950's-70's has rekindled the interested in molten salt-cooled reactors. The most recent form of this type of reactor, the Fluoride salt-cooled High-temperature nuclear Reactors (FHRs) is emerging as a leading reactor concept for the next generation nuclear reactors because it offers, among other benefits, a high degree of passive safety, high thermal efficiency, and low spent fuel. Since 2011, two three-year Integrated Research Projects (IRPs) were awarded by the U. S. Department of Energy (DoE) to the Massachusetts Institute of Technology (MIT) in partnership with the University of California, Berkeley and the University of Wisconsin-Madison to develop a 'path forward' for the development of the FHR. One primary challenge in these projects is to evaluate the corrosion and compatibility of structural alloys in the potential primary coolant of molten Li{sub 2}BeF{sub 4} (FLiBe) salt at 700 deg. C in a neutron irradiation environment. In support of structural materials development for the FHR, the static in-reactor corrosion tests of two promising alloys, nickel-based Hastelloy N{sup R} and 316 stainless steel, in {sup 7}Li-enriched molten FLiBe have been conducted in both metal-lined and graphite crucibles in the MIT nuclear research reactor in 2013. This work focuses on the microstructural characterization of the in-reactor corrosion tested alloys for understanding of their performance in high-temperature molten FLiBe salt. Our preliminary results indicate significant thermal effects, graphite effects, and irradiation effects on the alloys' performance in high-temperature molten FLiBe salt, particularly for the in-reactor corrosion tests. The neutron irradiation induces tritium generation in molten salt which could combine with fluorine ions to form corrosive HF to dramatically attack alloys. Additionally, the neutron irradiation introduces structural defects into alloys which probably alter alloys' corrosion resistivity to high temperature molten FLiBe salt. The effect of graphite on the alloys' corrosion in a nuclear reactor is also possibly different from the out-of-reactor corrosion. All these issues will be better understood by characterizing the microstructure of in-reactor corrosion tested alloy samples in our near future plans. The experimental data in this study provides valuable input for evaluating structural materials performance in the FHR. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

PWR Station Blackout and External Events Analysis for LWRS/RISMC Project

In the framework of the US-DOE Light Water Reactor Sustainability/ Risk-Informed Safety Margin Characterization (LWRS/RISMC) project, methods, data and tools are being developed with the ultimate scope of improving the safety and operation of the current US nuclear fleet. A set of Industrial Applications (IA) concerning the most relevant safety issues is defined for applying and testing the developed tool kits. The ultimate scope of IA no. 2 is a realistic simulation of natural external hazards that impose threat to a Nuclear Power Plant (NPP). This scope requires the development of a methodology and of a qualified set of tools able to perform advanced risk-informed safety analysis. In particular the methodology should be able to combine results from seismic, flooding and thermal-hydraulic (TH) deterministic calculations with dynamic Probabilistic Risk Assessment (PRA). A so-called Risk Informed Margin Management (RIMM) approach is being used to represent meaningful (i.e., realistic facility representation and simulation) event scenarios and consequences by linking probabilistic simulations to external events physics. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Uncertainty Quantification for External Events Analysis of LWRS/RISMC Project

In the framework of the US-DOE Light Water Reactor Sustainability/ Risk-Informed Safety Margin Characterization (LWRS/RISMC) project, methods, data and tools are being developed with the ultimate scope of improving the safety and operation of the current US nuclear fleet. A set of Industrial Applications (IA) concerning the most relevant safety issues is defined for applying and testing the developed toolkits. The ultimate scope of IA no. 2 is a realistic simulation of natural external hazards that impose threat to a Nuclear Power Plant (NPP). A methodology, which combines a qualified set of tools able to perform advanced risk-informed safety analysis, has been described in a previous paper. In this paper we present the details of the uncertainty quantification (UQ) for the system thermal-hydraulic code and how the UQ is integrated in the global external-events analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Acoustic Research under the Source Physics Experiment

The Source Physics Experiment series is a long-term research and development (R&D) effort under the U.S. Department of Energy’s National Nuclear Security Administration focused on improving the physical understanding of how chemical explosions generate seismoacoustic signals. Beginning in 2011, a series of subsurface chemical explosions in two different and highly contrasting geologies were conducted at the Nevada National Security Site in Nevada, USA with the objective of improving simulation and modeling approaches to explosion identification, yield estimation and other monitoring applications. The two executed phases of the series provide new explosion signature source data from a wide range of geophysical diagnostic equipment; recorded data from the test series is now openly available to the broader seismoacoustic community. This manuscript details the executed test series, deployed seismoacoustic networks, and summarizes major scientific achievements utilizing recorded signatures from the explosive tests.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2016

Heated Bundle Test in AREVA Inc's Fuel Cooling Test Facility

As part the Department of Energy (DOE) initiative for advanced nuclear fuel development, AREVA Inc is performing a series of flow and thermal mixing tests on a simulated, internally heated bundle of wire-wrapped fuel pins of the type commonly used in liquid-metal fast-breeder reactors at its Fuel Cooling Test Facility (FCTF) in Richland, Washington. The FCTF is part of the Product Development Test Facility (PDTF), which is a dedicated facility for advanced thermal-hydraulic testing of nuclear fuel assemblies. Initially developed to perform post Loss Of Coolant Accident (LOCA) testing, i.e., spray and reflood tests, this test facility has been adapted to accommodate the unique requirements of the heated bundle tests described here. The purpose of this test is to investigate and measure thermal distribution and pressure drop behavior for two configurations (non-deformed and deformed duct) of heated 61-pin, hexagonally packed fast reactor test assembly in a fully turbulent regime for generating Computational Fluid Dynamics (CFD) validation grade data. This test is performed in conjunction with high fidelity flow field measurement experiments and high fidelity CFD to provide a comprehensive understanding of the flow inside such a design. The present paper focuses on the heated bundle test performed at AREVA Inc. First, the test loop is described and then an overview of the test activities is presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U 3 Si 2 Modeling

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022~\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\ \noindent The purpose of this letter report is to highlight the multiscale modeling effort for \usi~fuel as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuel (ATF) High Impact Problem (HIP). This fiscal year two new models for \usi~fuel have been incorporated into the BISON fuel performance code~\cite{Williamson2012} based upon lower length scale simulations using the GRASS-SST rate theory code. The models are briefly described in the following sections.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Plan to Verify and Validate Multi-Hazard Risk-Informed Margin Management Methods and Tools

This report describes the development plan for verification and validation (V&V) of multihazard risk informed margin management (RIMM) methods and tools. The plan will include using new INL developed tool sets coupled with the new multi-partner External Hazards Experimental Group (EHEG) coordinated by the Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. The RISMC pathway is developing a suite of new tools and methods for advanced evaluation of facility risk. The RISMC toolkit, includes tools and methods focused on evaluation of risk from external hazards (e.g, seismic and flooding events), which have been shown to be dominant risk contributors in probabilistic risk assessments performed for operating nuclear power plants (NPPs). The external events activity (EEA) within the RISMC pathway is tasked with developing tools and methods focused on evaluation of multi-hazard external risk (e.g, seismic and flooding events). The EHEG will provide technical expertise and experimental large-scale testing data needed for development and validation of tools and methods in the RISMC toolkit for external hazard safety evaluations. The EEA within RISMC has two key elements: (1) an organizational and research framework provided by INL and (2) a coordinated group of university and national laboratory partners with the complementary expertise and experimental capabilities needed to conduct large-scale external hazard-focused experiments. This cooperative group will allow INL to leverage a range of existing capabilities to meet the unique needs of RISMC tool and method development. In addition, the capabilities of the EHEG could be used to address needs of other national laboratories. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high quality, large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases within RISMC. These databases will be used to validate the advanced external hazard tools and methods. The numerical tools under development in RISMC are used to evaluate the impacts that external hazards have on nuclear power plants. These analysis tools will be used to advanced current risk calculation processes and reduce uncertainty in these calculations. To have confidence in the predictive capability of these numerical tools it is important to verify and validate them. To verify and validate numerical tools the user must understand what physics is being used to represent the problem (i.e. constitutive models used to represent the soil and structure), the numerical solver approach used (i.e. finite element), what data already exists that could be used to validate the tools, and what data is needed for validation.

31274↗
DOE OSTI2020

Benchmark Modeling and Simulation of the FFTF LOFWOS Test #13 Using SAM

The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. In July 1986, a series of unprotected Loss of Flow Without Scram (LOFWOS) transients were performed in FFTF as part of the Passive Safety Testing (PST) program. The LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors left off, has been chosen as a benchmark case by IAEA to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of sodium fast reactor passive safety in an IAEA Coordinated Research Project (CRP), launched in October 2018. The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR safety analysis. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. To participate the IAEA CRP and enhance the SAM validation base for advanced reactor transient safety analysis, benchmark simulations of the FFTF LOFWOS Test #13 are performed using the SAM code. In this first phase of the validation effort, the thermal-hydraulic behavior of the reactor system is the focus and the reactor kinetics is not considered in the SAM FFTF model. Instead, the results of Argonne’s neutronics calculations are directly used, including the power shape of the active core region and the power history during the transient. The simulation results of FFTF at steady state agreed well with the measured data from the test. During the transient, reasonably good agreement were also obtained. Future work to improve the model will focus on introducing the reactivity predictions into the model, as well as better understanding or resolving the current discrepancies with the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compacts containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. Furthermore, in the case of krypton, all the coating layers are partly retentive and the available data did not allow the level of retention in individual layers to be determined, hence preventing derivation of any correction factors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Ground Motion Inputs for the Seismic Shake Table Test

Currently, spent nuclear fuel (SNF) is stored in on-site independent spent-fuel storage installations (ISFSIs) at seventythree (73) nuclear power plants (NPPs) in the US. Because a site for geologic repository for permanent disposal of SNF has not been constructed, the SNF will remain in dry storage significantly longer than planned. During this time, the ISFSIs, and potentially consolidated storage facilities, will experience earthquakes of different magnitudes. The dry storage systems are designed and licensed to withstand large seismic loads. When dry storage systems experience seismic loads, there are little data on the response of SNF assemblies contained within them. The Spent Fuel Waste Disposition (SFWD) program is planning to conduct a full-scale seismic shake table test to close the gap related to the seismic loads on the fuel assemblies in dry storage systems. This test will allow for quantifying the strains and accelerations on surrogate fuel assembly hardware and cladding during earthquakes of different magnitudes and frequency content. The main component of the test unit will be the full-scale NUHOMS 32 PTH2 dry storage canister. The canister will be loaded with three surrogate fuel assemblies and twenty-nine dummy assemblies. Two dry storage configurations will be tested – horizontal and vertical above-ground concrete overpacks. These configurations cover 91% of the current dry storage configurations. The major input into the shake table test are the seismic excitations or the earthquake ground motions – acceleration time histories in two horizontal and one vertical direction that will be applied to the shake table surface during the tests. The shake table surface represents the top of the concrete pad on which a dry storage system is placed. The goal of the ground motion task is to develop the ground motions that would be representative of the range of seismotectonic and other conditions that any site in the Western US (WUS) or Central Eastern US (CEUS) might entail. This task is challenging because of the large number of the ISFSI sites, variety of seismotectonic and site conditions, and effects that soil amplification, soil-structure interaction, and pad flexibility may have on the ground motions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Development of High Temperature Out-of-Pile Experiments for Testing Nuclear Thermal Propulsion Fuel Surrogates

Nuclear thermal propulsion (NTP) technology is designed to provide “in-space” vehicle propulsion by deriving thrust from a hydrogen working fluid that is heated from a nuclear reaction. Initial NTP fuel structures (primarily carbon/carbide based materials) developed by the Rover/NERVA program were designed to withstand temperatures exceeding 2400 K and remain intact after tens of firing cycles. Current programs responsible for development of this technology expect the fuels to withstand temperatures in excess of 2550K under a similar duty cycle. Experiments that subject fuels to these conditions are being developed to provide data and performance information for optimization and potential down-selection of the candidate materials. Initial tests will be designed for laboratory benchtop operation and will not expose surrogate specimens to the hydrogen working fluid or irradiation fields. The primary focus of these out-of-pile experiments will be to study thermal performance and thermal failure limits of the candidate specimen materials.

Howard, Richard↗