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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.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Nondestructive Testing Research and Development Efforts at Argonne National Laboratory—An Overview

The Nuclear Engineering (NE) Division of Argonne National Laboratory (ANL) has been involved for many years with research and development efforts associated with the nondestructive testing/evaluation (NDT/NDE) of critical components and structures in power generation (nuclear and fossil plants), aerospace, defense, homeland security, medical, and scientific applications. A wide range of NDT methods including electromagnetic, thermal, radiographic, acoustic and ultrasonic methods have been used for a wide range of material evaluation applications. Argonne’s expertise in analytical and numerical modeling of thermal, acoustic and electromagnetic interaction with complex media as well as advanced signal processing and data analysis have been utilized extensively for NDT applications. Three active areas of research, all in connection with NDT of materials and systems utilized in nuclear energy applications, are discussed in detail in this paper.

Eddy current testing↗
DOE OSTI2014

A new FPGA-based trigger/DAQ system for the MTV experiment at TRIUMF

A new FPGA based data acquisition (DAQ) and trigger combined system is developed for the MTV experiment (Mott Polarimetry for T-Violation Experiment). The MTV experiment measures backward Mott scattering events by detecting the incident and scattered tracks using a drift chamber. For the backward scattering probability is as small as less than 0.1%, efficient triggering system is required to select the desired events. An intelligent Level-2 trigger system reading hitting pattern together with its digital readout system are built in a custom FPGA system. In this paper, design overview and results from performance tests are described.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

IN-PILE MEASUREMENTS OF FUEL ROD DIMENSIONAL CHANGES UTILIZING THE TEST REACTOR LOOP PRESSURE FOR MOTION

Different types of materials are being considered for fuel, cladding, and structures in existing and advanced nuclear reactors. These materials can undergo significant dimensional and physical 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). The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach expensive. In addition, such techniques provide limited data, and handling may disturb the phenomena of interest. Therefore, in-pile detection of changes in geometry is sorely needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). This paper presents development 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. This test rig is being developed at the Idaho National Laboratory (INL), and will provide experimenters with a unique capability to measure dimensional changes associated with fuel and clad swelling, pellet-clad interaction, and crud buildup. This novel LVDT design uses the MTR’s loop pressure to drive the test specimen at a constant rate. This leads to a simplified construction, that is, the use of fewer components than other diameter gauge test rigs. This simplified design only requires three pressure boundary penetrations. Further, a key component of this test rig is a leak tight hydraulic ram capable of actuating under low differential pressure, and is able to survive the high irradiation tests at Pressurized Water Reactor (PWR) conditions. The sensor has been proven on a benchtop scale, and shows repeatability and reliability. Measurement accuracies of ±3 µm for diameter and ±90 µm for length have been achieved.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

IER 516 - Zirconium Test Assembly (ZTA) Preliminary Experimental Design [Slides]

Zirconium present in large quantities in criticality safety applications such as fuel manufacturing and processing facilities, and in commercial thermal nuclear reactors. ICSBEP lacks benchmarks optimized for zirconium sensitivity to validate zirconium evaluations. Evaluators lack isotopic differential nuclear data in regions of high importance to fission systems.

K-PSO 2.0↗
DOE OSTI2019

NSUF Research Reactor Fitness Workshop Results

The Nuclear Science User Facilities has been working with the National Organization of Test, Research, and Training Reactors to collect input from the US university reactor community on their current and future needs. The effort started with a web survey in March to collect initial data. NSUF held a follow-up workshop in July at the Center for Advanced Energy Studies in Idaho Falls. We identified four areas of interest in the university research reactor community that can affect future operations and sustainability of the reactors: capital infrastructure, regulatory burden, staffing and knowledge transfer, and utilization and relevancy. Members of the community presented challenges and best practices from their facilities. Discussions were held to create and distill the list of challenges for each area. A working group was formed for each area to continue the discussion and prioritize the challenges. NSUF and the working groups intend to create a final recommendations report by the end of the year. This panel session will discuss the workshop and summarize the discussion in the four topic areas. The audience is encouraged to add to the discussion.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Performance of an RPM based on Gd-lined plastic scintillator for neutron and gamma detection [ANIMMA--2015-IO-372]

A Radiation Portal Monitor (RPM) was developed by the Istituto Nazionale di Fisica Nucleare (INFN) and Ansaldo Nucleare (ANN) within the FP7 SCINTILLA European project. The system was designed to detect both gamma and neutron radiation with a single technology. It is conceived to monitor vehicle and cargo containers in transits across borders or ports, to find radioactive elements and to avoid illegal trafficking of strategic nuclear materials. The system is based on a {sup 3}He-free neutron detection technology using plastic scintillators coupled to Gadolinium to detect and discriminate gamma from neutron signals. During the 3 years of the SCINTILLA project the construction and test of the first two prototypes drove the definition of the final layout of a full RPM system consisting of two twin pillars as a portal for vehicle and cargo container scan. A custom System Control Software (SCS) manages the electronics of the RPM, the ancillary devices and the data analysis. The combination of the detector layout and of the software functionalities enables both to distinguish neutrons and gammas and to identify the energy range of a detected gamma source. The system was initially characterized via static tests with gamma and neutron sources in the INFN laboratory. These measurements were used to calibrate the detector, evaluate the response of the single pillars as well as of the full system, and optimize the RPM configuration and discrimination algorithm. During this phase, specific tests were performed to study the stability over time of the system, monitoring the measured the neutron and gamma count rates over periods of several weeks. The results allow us to demonstrate the reliability and robustness of the RPM. In a second time the RPM performance was studied via dynamic tests performed during the SCINTILLA test and benchmark campaigns. These measurements took place in the JRC ITRAP+10 facility at Ispra (Varese-Italy). The laboratory is equipped with an experimental set-up for dynamic tests of multiple systems according to international standards. The performed measurements utilized radioactive sources with activities selected according to ANSI and IEC standards to test the detector alarm performances in terms of gamma and neutron response, sensitivity to high gamma fields, sensitivity to moderated neutron sources as well as false alarm rates (FAR). In addition, the RPM was tested in challenging configurations exceeding the requirements set by international standards to determine the real limits of the system. The results obtained during these campaigns demonstrated that the system detection efficiency is not only compliant to international standards for its category, but often exceeds them, demonstrating the validity of the chosen technology and of the implemented layout. The positive performance also showed the effectiveness of the SCS and of its functionalities. To further demonstrate the system capabilities, a test in a real-life environment of the RPM is planned to happen in a near future by installing the detectors in a seaport. In this presentation I will give an overview of the RPM characteristics, of its performances as determined in the test campaign mentioned above and of future plans, to demonstrate how this technology can be an effective choice for the realization of {sup 3}He-free RPM detectors. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Radioxenon spiked air

Four of the radioactive xenon isotopes ( 131m Xe, 133m Xe, 133 Xe and 135 Xe) with half-lives ranging from 9 h to 12 days are produced from nuclear fission and can be detected from days to weeks following their production and release. Being inert gases, they are readily transported through the atmosphere. Sources for release of radioactive xenon isotopes include operating nuclear reactors via leaks in fuel rods, medical isotope production facilities, and nuclear weapons' detonations. They are not normally released from fuel reprocessing due to the short half-lives. The Comprehensive Nuclear-Test-Ban Treaty has led to creation of the International Monitoring System. The International Monitoring System, when fully implemented, will consist of one component with 40 stations monitoring radioactive xenon around the globe. Monitoring these radioactive xenon isotopes is important to the Comprehensive Nuclear-Test-Ban Treaty in determining whether a seismically detected event is or is not a nuclear detonation. A variety of radioactive xenon quality control check standards, quantitatively spiked into various gas matrices, could be used to demonstrate that these stations are operating on the same basis in order to bolster defensibility of data across the International Monitoring System. This study focuses on Idaho National Laboratory's capability to produce three of the xenon isotopes in pure form and the use of the four xenon isotopes in various combinations to produce radioactive xenon spiked air samples that could be subsequently distributed to participating facilities.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Creep-fatigue behavior of alloy 617 - 18581

Alloy 617 is the leading candidate material for the intermediate heat exchanger for the very high temperature reactor (VHTR). It is thought that creep-fatigue resistance is likely to be the life-limiting material property for this application. As part of evaluating the behavior of this material in the expected service conditions, creep fatigue testing of Alloy 617 was performed at 850 and 950 deg. C for several strain ranges and hold times. Creep-fatigue degradation was observed, characterized by reduced cycles to failure when a hold time was added at peak tensile strain, compared to fatigue values. At 850 deg. C, increasing the tensile hold duration continued to degrade the creep fatigue resistance, at least to the investigated strain controlled hold time of up to 60 minutes at the 0.3% strain range and 240 minutes at the 1.0% strain range. At 950 deg. C, the creep fatigue cycles to failure are not further reduced with increasing hold duration, indicating saturation occurs at relatively short hold times. The creep and fatigue damage fractions have been calculated using the ASME methodology and plotted on a creep fatigue interaction D diagram. The impact of selecting different parameters on calculation of the creep damage fraction has been evaluated and will be discussed in this paper. Additional damage fractions from creep fatigue tests at 800 and 1000 deg. C on a different heat of Alloy 617 have been determined and are plotted on the D diagram along with results of the current testing. The combined data represent a comprehensive evaluation of Alloy 617 behavior under the ASME methodology and result in a consistent creep-fatigue damage envelope. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Validation of Modern Nuclear Data Processing in SCALE

The nuclear data (ND) community is continuously developing more accurate, diversified, and comprehensive data for radiation transport modeling to support the nuclear science community. As these community efforts progress, it is crucial that ND processing tools like AMPX (used for SCALE [1] ND) also be developed in parallel to incorporate these new data into transport codes and actually deliver those data to end users. AMPX is a mature, well-tested code that was developed by many people at Oak Ridge National Laboratory (ORNL) over the course of the past few decades. A large portion of the AMPX codebase, however, was outdated, difficult to maintain, and incompatible with modern code development tools. Some of the most important parts of the AMPX code have now been replaced with modern C++ code that can be maintained more cost-effectively and can be tested more rigorously.

AMPX↗
DOE OSTI2015

NEET-AMM Final Technical Report on Laser Direct Manufacturing (LDM) for Nuclear Power Components

Final technical report summarizes the program progress and technical accomplishments of the Laser Direct Manufacturing (LDM) for Nuclear Power Components project. A series of experiments varying build process parameters (scan speed and laser power) were conducted at the outset to establish the optimal build conditions for each of the alloys. Fabrication was completed in collaboration with Quad City Manufacturing Laboratory (QCML). The density of all sample specimens was measured and compared to literature values. Optimal build process conditions giving fabricated part densities close to literature values were chosen for making mechanical test coupons. Test coupons whose principal axis is on the x-y plane (perpendicular to build direction) and on the z plane (parallel to build direction) were built and tested as part of the experimental build matrix to understand the impact of the anisotropic nature of the process.. Investigations are described 316L SS, Inconel 600, 718 and 800 and oxide dispersion strengthed 316L SS (Yttria) alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Validation of HELIOS for ATR Core Follow Analyses

This work summarizes the validation analyses for the HELIOS code to support core design and safety assurance calculations of the Advanced Test Reactor (ATR). Past and current core safety assurance is performed by the PDQ-7 diffusion code; a state of the art reactor physics simulation tool from the nuclear industry’s earlier days. Over the past twenty years, improvements in computational speed have enabled the use of modern neutron transport methodologies to replace the role of diffusion theory for simulation of complex systems, such as the ATR. More exact methodologies have enabled a paradigm-shift away from highly tuned codes that force compliance with a bounding safety envelope, and towards codes regularly validated against routine measurements. To validate HELIOS, the 16 ATR operational cycles from late-2009 to present were modeled. The computed power distribution was compared against data collected by the ATR’s on-line power surveillance system. It was found that the ATR’s lobe-powers could be determined with ±10% accuracy. Also, the ATR’s cold startup shim configuration for each of these 16 cycles was estimated and compared against the reported critical position from the reactor log-book. HELIOS successfully predicted criticality within the tolerance set by the ATR startup procedure for 13 out of the 16 cycles. This is compared to 12 times for PDQ (without empirical adjustment). These findings, as well as other insights discussed in this report, suggest that HELIOS is highly suited for replacing PDQ for core safety assurance of the ATR. Furthermore, a modern verification and validation framework has been established that allows reactor and fuel performance data to be computed with a known degree of accuracy and stated uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Updates from the Covariance Session Committee [Slides]

The CSEWG covariance session template paper was submitted on June 8, 2021 to journal. CSEWG will not publish adjusted VIII.1, but there is interest in adjustment results from the community. Other focus points for the release for VIII.1 include stringent testing of covariances before their release and missing covariances.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2017

Sandia National Laboratories, Tonopah Test Range Fire Control Bunker (Building 09-51): Photographs and Written Historical and Descriptive Data

The Fire Control Bunker (Building 09-51) is a contributing element to the Sandia National Laboratories (SNL) Tonopah Test Range (TTR) Historic District. The SNL TTR Historic District played a significant role in U.S. Cold War history in the areas of stockpile surveillance and non-nuclear field testing of nuclear weapons design. The district covers approximately 179,200 acres and illustrates Cold War development testing of nuclear weapons components and systems. This report includes historical information, architectural information, sources of information, project information, maps, blueprints, and photographs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

MPACT Verification and Validation Manual (rev 1)

The MPACT code provides the capability to calculate the steady-state and transient reactor core neutron distribution for operating nuclear power plant (PWR) conditions across multiple fuel cycles. As required in the CASL SQA plan, it is the responsibility of UM and ORNL as joint owners of MPACT to ensure that verification and validation activities are performed and documented in a V&V manual with supporting publications and CASL technical reports which can be distributed for reference and distribution within VERA. This document provides the first revision of the MPACT verification and validation (V&V) manual and describes the current state of MPACT V&V and updates the framework for future MPACT V&V activities. During FY16, MPACT V&V activities in the areas of code verification have continued to mature and work has progressed on code error evaluation using the Method of Manufactured Solutions (MMS). Several specific tasks have also been completed to improve MPACT code verification to include tasks to improve the unit and regression test coverage. A joint UM/ORNL task was completed to develop standardize coding standards and workflow for the continued collaborative development of MPACT. During the past year, MPACT validation work has continued in both the areas of measured data from critical experiments as well as measured data from operating nuclear power plants. Preliminary work was also performed in FY16 on the use of post-irradiation examinations for evaluation and validation of the isotopic depletion capability in MPACT. Based on the results of V&V efforts this past year, the confidence level has increased in the ability of MPACT to model an operational Pressurized Water Reactor. Areas were identified which require increased emphasis during FY17 to include updating all the V&V problems with the new 51-group library, the completion of cycle 2 of the BEAVRS benchmark, initial validation for BWR applications, the addition of Uncertainty Quantification results into the benchmark results, and the completion of specific validation tasks for the CASL Challenge problems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Application of photon Doppler velocimetry to direct impact Hopkinson pressure bars

Direct impact Hopkinson pressure bar systems offer many potential advantages over split Hopkinson pressure bars, including access to higher strain rates, higher strains for equivalent striker velocity and system length, lower dispersion, and faster achievement of force equilibrium. Currently, these advantages are gained at the expense of all information about the striker impacted specimen face, preventing the experimental determination of force equilibrium, and requiring approximations to be made on the sample deformation history. In this paper, we discuss an experimental method and complementary data analysis for using photon Doppler velocimetry to measure surface velocities of the striker and output bars in a direct impact bar experiment, allowing similar data to be recorded as in a split bar system. We discuss extracting velocity and force measurements, and the precision of measurements. Results obtained using the technique are compared to equivalent split bar tests, showing improved stress measurements for the lowest and highest strains in fully dense metals, and improvement for all strains in slow and non-equilibrating materials.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Measurement of the proton spin structure at long distances

Measuring the spin structure of protons and neutrons tests our understanding of how they arise from quarks and gluons, the fundamental building blocks of nuclear matter. At long distances, the coupling constant of the strong interaction becomes large, requiring non-perturbative methods to calculate quantum chromodynamics processes, such as lattice gauge theory or effective field theories. In this work, we report proton spin structure measurements from scattering a polarized electron beam off polarized protons. The spin-dependent cross-sections were measured at large distances, corresponding to the region of low momentum transfer squared between 0.012 and 1.0GeV 2 . This kinematic range provides unique tests of chiral effective field theory predictions. Our results show that a complete description of the nucleon spin remains elusive, and call for further theoretical works, for example, in lattice quantum chromodynamics. Finally, our data extrapolated to the photon point agree with the Gerasimov–Drell–Hearn sum rule, a fundamental prediction of quantum field theory that relates the anomalous magnetic moment of the proton to its integrated spin-dependent cross-sections.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2020

AFIP6-MkII and RERTR-12 Porosity Data Collection and Analysis for Modeling and Simulation

Gathering data for the improvement of nuclear fuel modeling and simulation efforts is the primary driver for this work. Mechanistic models allow for a better understanding of the material on a micro- and macrostructural level while saving time and money over traditional experiment efforts. Historically, summarized data and correlations are the inputs for empirical material models and model validation. When improving these models for nuclear fuels with experimental results, there is a lack of reliable data readily available. Experiments - RERTR-12 and AFIP6-MkII - were conducted to understand the irradiation behavior of metallic U-10Mo monolithic fuels for use in extreme reactor environments such as research reactors like the Advanced Test Reactor (ATR) or the High Flux Isotope Reactor (HFIR). Microstructural characteristics of fission gas pores (FGP) in each experiment are collected using an automated image analysis technique developed at the University of Florida and presented here. A series of statistical tests are performed to explore the reliability of the results, as well as understand where the data is lacking and what future data collection is necessary to provide sufficient information to assist modeling efforts. The focus is on the porosity, pore size, and eccentricity of FGPs formed during irradiation in three AFIP6-MkII samples and one RERTR-12 sample. From the analysis, it is clear there are substantial impacts of fission density on the pore structure, but there also exist also underlying connections between each sample and the behavior observed in the pores. Further analyses of the pre- and post-irradiation microstructure are needed to improve the understanding of these connections. An early method for microstructural data analysis is presented within and is currently being expanded to include other microstructure data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal Neutron Scattering Evaluation Methods Development at ORNL [Slides]

The goal of the project was to develop a procedure to generate covariance matrix for S( α,β ) data that incorporated both computational simulations and experimental data. Experimental data and computer simulation fit is achieved using the Unified Monte Carlo (UMC) method. The framework is material & simulation method independent, and this is demonstrated using light water. A methodology was developed for generalized generation of covariances for thermal scattering files. The framework was subsequently tested on light water by generating covariance of pDOS with encouraging results. Preliminary results suggest temperature effects of polyethylene may warrant further investigation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗