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 1,153 records · Page 64
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

In-pile measurements of fuel rod dimensional changes utilizing the test reactor loop pressure for motion - 359

Different types of materials are being considered for fuel, cladding, and structures in existing and advanced nuclear reactors. These materials can undergo significant dimensional changes during irradiation. Currently in the US, such changes are measured with the 'cook and look' method (i.e. repeatedly irradiating a specimen for a specified period of time, and then removing it from the reactor for evaluation) - which leads to costly and time consuming experiments. In addition, such techniques provide limited data, and handling may disturb the phenomena of interest. Therefore, in-pile detection of geometric changes is needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). The developmental results of an advanced Linear Variable Differential Transformer (LVDT)-based test rig capable of detecting real-time changes in diameter or length of fuel rods or other material samples during irradiation in US MTRs are presented. The LVDT is being developed at the Idaho National Laboratory (INL), and will provide experimenters with the capability to measure dimensional changes associated with fuel and clad swelling, pellet-clad interaction, and crud buildup. The LVDT design utilizes the MTR's loop pressure to drive the test specimen at a constant rate. This leads to a simplified construction which only requires three pressure boundary penetrations. Further, the LVDT pressure boundary is a leak tight hydraulic ram capable of actuating under low differential pressure (e.g. 1 psid), and is able to survive the high irradiation tests at Pressurized Water Reactor (PWR) conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

AGR-3/4 Final Data Qualification Report for ATR Cycles 151A through 155B-1

This report provides the qualification status of experimental data for the entire Advanced Gas Reactor 3/4 (AGR 3/4) fuel irradiation. AGR-3/4 is the third in a series of planned irradiation experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the AGR Fuel Development and Qualification Program, which supports development of the advanced reactor technology under the INL ART Technology Development Office (TDO). The main objective of AGR-3/4 irradiation is to provide a known source of fission products for subsequent transport through compact matrix and structural graphite materials due to the presence of designed-to-fail fuel particles. Full power irradiation of the AGR 3/4 test began on December 14, 2011 (ATR Cycle 151A), and was completed on April 12, 2014 (end of ATR Cycle 155B) after 369.1 effective full power days of irradiation. The AGR-3/4 test was in the reactor core for eight of the ten ATR cycles between 151A and 155B. During the unplanned outage cycle, 153A, the experiment was removed from the ATR northeast flux trap (NEFT) location and stored in the ATR canal. This was to prevent overheating of fuel compacts due to higher than normal ATR power during the subsequent Powered Axial Locator Mechanism cycle, 153B. The AGR 3/4 test was inserted back into the ATR NEFT location during the outage of ATR Cycle 154A on April 26, 2013. Therefore, the AGR-3/4 irradiation data received during these 2 cycles (153A and 153B) are irrelevant and their qualification status isnot included in this report. Additionally, during ATR Cycle 152A the ATR core ran at low power for a short enough duration that the irradiation data are not used for physics and thermal calculations. However, the qualification status of irradiation data for this cycle is still covered in this report. As a result, this report includes data from 8 ATR Cycles: 151A, 151B, 152A, 152B, 154A, 154B, 155A, and 155B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). The AGR 3/4 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates, pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates, release to birth rate ratios [R/Bs], and particle failure counts) for each of the twelve capsules in the AGR 3/4 experiment. During Outage Cycle 155A, fourteen flow meters were installed downstream from fourteen FPMS monitors to measure flows from the monitors; qualification status of these data are also included in the report. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) composed of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. For ATR Cycles 151A through 154B, the DRC convened on February 12, 2014, reviewed the data acquisition process, and considered whether the data met the requirements for data collection as specified in QA approved INL ART TDO data collection plans. The DRC also examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report. The qualification status of AGR-3/4 irradiation data during the first six cycles were previously reported in INL/EXT-14-31186 document. This report presents data qualification status for the entire AGR-3/4 irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Summary of the Microstructural Characterization Results for the A709 Commercial Heats

To improve high temperature strength for advanced nuclear reactors, it is necessary to qualify additional structural materials for introduction into the American Society of Mechanical Engineers (AMSE) Boiler and Pressure Vessel Code (BPVC). Specifically, Alloy 709 was identified as a material for Sodium-cooled Fast Reactors to enhance time-dependent mechanical properties compared to 316H stainless steel, a code qualified high temperature alloy. Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) performed initial microstructure and mechanical property testing on a third, commercial heat of Alloy 709 stainless steel. The Alloy 709 plate was produced by Allegheny Technology Inc. (ATI) and in the solution annealed condition. The chemical composition was within the limits for an American Society of Testing and Materials (ASTM) A213 specification for seamless tube for material UNS S31025. Grain size was measured as an average ASTM size of 6.5, which was larger than the ASTM 7 required according to ASTM A213 for seamless tube. The chemical composition was within the limits for an ASTM A213, and the room temperature tensile tests and hardness data met ASTM A213 specifications. Creep-fatigue results showed similar behavior to previous heats of Alloy 709.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

An ECT/ERT dual-modality sensor for oil-water two-phase flow measurement

This paper presents a new sensor for ECT/ERT dual-modality system which can simultaneously obtain the permittivity and conductivity of the materials in the pipeline. Quasi-static electromagnetic fields are produced by the inner electrodes array sensor of electrical capacitance tomography (ECT) system. The results of simulation show that the data of permittivity and conductivity can be simultaneously obtained from the same measurement electrode and the fusion of two kinds of data may improve the quality of the reconstructed images. For uniform oil-water mixtures, the performance of designed dual-modality sensor for measuring the various oil fractions has been tested on representative data and the results of experiments show that the designed sensor broadens the measurement range compared to single modality.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS2012

Compact Fuel Element Environment Test

Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

Bradley, D. E.↗
DOE OSTI2022

Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Separate Effects Miniature Fuel Irradiation Testing to Support Fuel Safety

The recent interest in advanced fuel forms that offer increased thermal conductivity and uranium density, as well as modifications (e.g., use of dopants or additives) to traditional UO2 fuel, necessitates a vehicle for performing high-throughput irradiation testing of these fuels before they can be safely deployed. This interest, combined with the recent decommissioning of the Halden Reactor, places a strong burden on the remaining test reactors to provide the scientific data needed to advance these fuels toward commercialization. Oak Ridge National Laboratory is actively working to alleviate some of this burden by commissioning a new “MiniFuel” test facility that will provide a vehicle for performing accelerated separate-effects irradiation testing of sub-sized nuclear fuel specimens. The small size of the fuel allows for a flexible irradiation vehicle capable of exploring a wide range of fuel performance variables within a reasonable time and cost. This paper describes the experiment facility, and the initial irradiation experiments being performed including uranium nitride (UN) kernels, coated tristructural-isotropic (TRISO) UN particles, doped and un-doped (reference) UO2 disks, U3Si2 disks, and fully ceramic microencapsulated (FCM) pellets with UN kernels and a SiC matrix. A focus is placed on the performance of miniature FCM pellets, which are being considered as the ultimate accident tolerant fuel for light water reactors. The experiments described in this paper will provide basic scientific data (e.g., swelling, fission gas release) regarding irradiation performance under well-characterized conditions.

Petrie, Christian↗
DOE OSTI2018

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

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

Baseline Characterization Database Verification Report – PCEA Billet XPC01D3-36

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) program. This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Data Management and Analysis System (NDMAS), and from that database will help populate the handbook of property data available to member nations of the Generation IV International Forum (GIF). The transfer of this data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for PCEA billet XPC01D3-36 and facilitates the release of the associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Benchmark of Spacer Grid Models used in Nuclear Power Plants

The fuel bundles of both PWR and BWR reactors require spacers to maintain the integrity and spacing of the slender fuel rods. These spacers can provide benefits such as increased mixing and turbulence, which increases heat distribution, leading to better Departure from Nucleate Boiling (DNB) conditions in PWRs. They also enhance the critical power ratios by increasing the thickness of the liquid layer in BWRs. However, the spacers also cause a pressure drop which must be engineered into the performance and operation. The flow within a rod bundle involves a number of complicated factors, such as cross flow, (de)entrainment of liquid layer, bubble or droplet break up/join. This requires specialized 3 dimensional computational fluid dynamics for the complete flow, however, most of the newly developed models have been accomplished by industry, so they are proprietary and in addition require extensive programming and cost to implement. In the public domain, a number of 1 dimensional equations have been developed that could benefit in initial design and decision making, if the equations are valid in the new and different applications the models are used. In 2002, an OECD/NRC BFBT benchmark workshop determined the variables used in the development of models, to predict flow within nuclear fuel bundles, had not been updated with technological advances to replace general theory and characteristics. If those properties could be identified and modeled, they could be applied to a wider range of conditions. This study utilizes NUPEC BFBT tests' measured data, on an average spacer over a range of pressure and flows, which was determined to be the most detailed measured data available. The focus is to describe the mathematical models of spacer grids and validate them with experimental results. This study consists of compiling and comparing available and established single phase equations and models computations. While these models may not have the desired degree of accuracy for final decisions, they may be accurate enough for the initial phases of design. No conclusion on a universal equation that fits all boundary conditions accurately or detailed effect of properties was determined. However, by using the BFBT experimental data, boundary conditions where the spacer grid models can be useful were identified. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Glass-contact refractory of the nuclear waste vitrification melters in the United States: a review of corrosion data and melter life

The performance of the refractory lining in glass melters used for nuclear waste vitrification is critical to the melter reliability for long-term continuous operation. Monofrax® K-3, a high Cr 2 O 3 fused cast refractory material, has been widely used to build the liners of nuclear waste glass melters in the United States. Corrosion behaviour of Monofrax® K-3 refractory has been evaluated based on crucible-scale testing, inspection of the refractory components following scaled melter testing, and inspections of the Defense Waste Processing Facility (DWPF) melter refractory after service. The literature generally consists of empirical models based on short-term testing to describe refractory corrosion dependence on glass composition. Corrosion data from tests with longer testing times, at various temperatures, in the presence of molten salts, and with different redox reactions in the plenum atmosphere exist, may be insufficient to provide accurate refractory service life estimates. Additionally, the corrosion data collected under actual and scaled melter operating conditions are limited. Recommendations to achieve more direct correlation between the laboratory refractory corrosion data predictions and the observed melter service life are discussed to allow for more accurate predictions of the useful life of melter refractory linings.

Jin, Tongan↗
DOE OSTI2015

Annex C. Data Libraries for IAEA Nuclear Security Assessment Methodologies (NUSAM)

Data Libraries are a collection of experimental data for PP elements based on specific defeat mechanisms. The data can de related to probability of detection, assessment, or barrier delay times. The data libraries can be a manually tabulated list or an electronic menu values enbedded in a stand-alone database or in an assessment software program such as SAVI or EASI. Data library values are usually based on years of historical testing data for many forms of common and sometimes unique PPS systems. Most useful testing libraries include data from the most basic alarm sensors, doors, walls, and barriers to very sophisticated PP elements. Each PP element is tested using increasing levels of defeat techniques from hand tools, power tools, breaching tools, explosives, and vehicles, as applicable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23) Appendix A: Full-Length Rod Heat Treatments

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase 1 DE activities related to full-length rod heat treatments (FHT) applied to selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Analysis of Localized Deformation Processes in Highly Irradiated Austenitic Stainless Steel through In Situ Techniques

This report describes new experimental results obtained during in situ mechanical tests with neutron-irradiated (10.7 dpa) austenitic 304L steel specimens. The in situ tensile tests inside a scanning electron microscope (SEM) were accompanied by modern high-resolution electron backscatter diffraction (HR-EBSD) analysis. The HR-EBSD approach allowed for investigating and quantifying internal stresses and dislocation structures in nondeformed and deformed steels, materials used in light water reactor internal components. The literature regarding the HR-EBSD analysis of metals and alloys is very limited, and practically no publications on the HR-EBSD analysis coupled with in situ tests are available. Experimental work, performed at Oak Ridge National Laboratory’s Low Activation Materials Development and Analysis facility, is detailed in this report. Section 1 briefly describes strain and stress localization processes and underlines their importance and role in irradiated materials, materials used in nuclear power plants. The section demonstrates that very limited quantitative data exist regarding localized stresses and strains in the irradiated materials at the meso-level (i.e., inside the grain, for a group of interacting grains, or for channel–grain boundary [GB] interaction) whereas local stresses at GBs are key contributors to the stress corrosion crack initiation.Section 2 documents the key methods used to conduct the experimental work: the SEM/EBSD system, miniature tensile frame, and irradiated material. Irradiated specimen manufacturing is described, as this option may be of special interest for other projects. Section 3 provides details of the HR-EBSD in situ tests (e.g., tensile curves, geometry, dimensions of regions of interest, load, and deformation steps). Section 4 discusses the results related to the evolution of geometrically necessary dislocations (GNDs) and their density, spatial, and statistical distributions.Section 5 focuses on the HR-EBSD analysis with respect to a channel–GB interaction. Two aspects are discussed in detail: mapping and analyzing the GND fields before and after loading, and the possibility of investigating local stresses and elastic strains at the channel–GB interaction. The present work results are beneficial for understanding strain and stress localization processes leading to crack initiation, for analyzing the material after long-term in-service life, and for developing predictive models. Section 6 summarizes the work performed and discusses future activities within this research direction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

The trigger and data acquisition for the NEMO-Phase 2 tower

In the framework of the Phase 2 of the NEMO neutrino telescope project, a tower with 32 optical modules is being operated since march 2013. A new scalable Trigger and Data Acquisition System (TriDAS) has been developed and extensively tested with the data from this tower. Adopting the all-data-to-shore concept, the NEMO TriDAS is optimized to deal with a continuous data-stream from off-shore to on-shore with a large bandwidth. The TriDAS consists of four computing layers: (i) data aggregation of isochronal hits from all optical modules; (ii) data filtering by means of concurrent trigger algorithms; (iii) composition of the filtered events into post-trigger files; (iv) persistent data storage. The TriDAS implementation is reported together with a review of dedicated on-line monitoring tools.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

NCSP Analytical Methods Subtask 3, AMPX Development and Maintenance, and NCSP Nuclear Data Subtask 6, SAMMY Modernization

SAMMY 1 is a production code that still contains a lot of old-style FORTRAN (F77) code. While the code is widely used and performs well, the legacy code base makes it difficult to maintain and to add new features. However, because it is a production code, it cannot lose any functionality while being modernized. In addition, new features need to be added to support the changing needs of the evaluators to analyze data obtained with new equipment and at higher resolution. In addition, as the data are analyzed with higher fidelity and new models within SAMMY, the processing codes and formats must keep pace. This will ensure that any new SAMMY evaluation can be processed by AMPX and can be tested and used in transport codes like SCALE 2 for maximum benefit of the Nuclear Criticality Safety Program (NCSP) mission.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗