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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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DOE OSTI2024

NEA HTTR LOFC Project Test#3 Benchmark Results

In the second half of FY23, the High Temperature Engineering Test Reactor (HTTR) Loss Of Forced Cooling (LOFC)#3 data for the 9 MW test case with Vessel Cooling System (VCS) off were made available through the Nuclear Energy Agency (NEA) LOFC project framework; the neutronic model developed for the initial test (LOFC#1) achieved a satisfactory level of maturity, demonstrating its accuracy in predicting power evolution and core re-criticality, but LOFC#3 should be used primarily to investigate thermal hydraulic phenomena, as the reactor was shut down prematurely due to overheating in the upper reactor components, which prevented re-criticality; this report focuses on advancing the HTTR thermal hydraulic model to accurately simulate the LOFC#3 scenario, including simulating the LOFC#3 benchmark and generating the corresponding benchmark specifications, aiming to ensure consistency across participant models and provide essential data for future participants, including private industry stakeholders seeking to validate their computational tools.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

FAUST - Benchmark and Validation Framework [Slides]

This presentation discusses Python packages for nuclear data applications and for benchmarking. It also discusses the goals and objectives for FAUST which include providing input and output processing for different calculation codes, allowing for exchanging results between different applications and codes, running benchmarks and processing the results, automating and simplify plot and report generation, and providing a basis for developing applications useful for nuclear data evaluators. The presentation also discusses how the packages can be used both inside and outside LANL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Development of Data Acquisition System for nuclear thermal hydraulic out-of-pile facility using the graphical programming methods

This paper describes the design of data acquisition system (DAQ) that is connected to a PC and development of a feedback control system that maintains the coolant temperature of the process at a desired set point using a digital controller system based on the graphical programming language. The paper will provide details about the data acquisition unit, shows the implementation of the controller, and present test results. (authors)

42 ENGINEERING↗
DOE OSTI2017

Weapon Physicist Declassifies Rescued Nuclear Test Films

The U.S. conducted 210 atmospheric nuclear tests between 1945 and 1962, with multiple cameras capturing each event at around 2,400 frames per second. But in the decades since, around 10,000 of these films sat idle, scattered across the country in high-security vaults. Not only were they gathering dust, the film material itself was slowly decomposing, bringing the data they contained to the brink of being lost forever. For the past five years, Lawrence Livermore National Laboratory (LLNL) weapon physicist Greg Spriggs and a crack team of film experts, archivists and software developers have been on a mission to hunt down, scan, reanalyze and declassify these decomposing films. The goals are to preserve the films’ content before it’s lost forever, and provide better data to the post-testing-era scientists who use computer codes to help certify that the aging U.S. nuclear deterrent remains safe, secure and effective.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2012

Development of autonomous gamma dose logger for environmental monitoring

Continuous monitoring and archiving of background radiation levels in and around the nuclear installation is essential and the data would be of immense use during analysis of any untoward incidents. A portable Geiger Muller detector based autonomous gamma dose logger (AGDL) for environmental monitoring is indigenously designed and developed. The system operations are controlled by microcontroller (AT89S52) and the main features of the system are software data acquisition, real time LCD display of radiation level, data archiving at removable compact flash card. The complete system operates on 12 V battery backed up by solar panel and hence the system is totally portable and ideal for field use. The system has been calibrated with Co-60 source (8.1 MBq) at various source-detector distances. The system is field tested and performance evaluation is carried out. This paper covers the design considerations of the hardware, software architecture of the system along with details of the front-end operation of the autonomous gamma dose logger and the data file formats. The data gathered during field testing and inter comparison with GammaTRACER are also presented in the paper. AGDL has shown excellent correlation with energy fluence monitor tuned to identify {sup 41}Ar, proving its utility for real-time plume tracking and source term estimation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2010

Testing of Small Graphite Samples for Nuclear Qualification

Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY22) Appendix E: Mechanical Testing

As a part of the 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 appendix documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Radiation effects in concrete for nuclear power plants, Part II: Perspective from micromechanical modeling

The need to understand and characterize the effects of neutron irradiation on concrete has become urgent because of the possible extension of service life of many nuclear power generating stations. Current knowledge is primarily based on a collection of data obtained in test reactors. These results are inherently difficult to interpret because materials and testing conditions are inconsistent. A micromechanical approach based on the Hashin composite sphere model is presented to derive a first-order separation of the effects of radiation on cement paste and aggregate, and, also, on their interaction. Although the scarcity of available data limits the validation of the model, it appears that, without negating a possible gamma-ray induced effect, the neutron-induced damage and swelling of aggregate plays a predominant role on the overall concrete expansion and the damage of the cement paste. Finally, the radiation-induced volumetric expansion (RIVE) effects can also be aided by temperature elevation and shrinkage in the cement paste.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

AGR-2 Final Data Qualification Report for U.S. Capsules - ATR Cycles 147A Through 154B

This report provides the data qualification status of AGR-2 fuel irradiation experimental data in four U.S. capsules from all 15 Advanced Test Reactor (ATR) Cycles 147A, 148A, 148B, 149A, 149B, 150A, 150B, 151A, 151B, 152A, 152B, 153A, 153B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Thus, this report covers data qualification status for the entire AGR-2 irradiation and will replace four previously issued AGR-2 data qualification reports (e.g., INL/EXT-11-22798, INL/EXT-12-26184, INL/EXT-13-29701, and INL/EXT-13-30750). During AGR-2 irradiation, two cycles, 152A and 153A, occurred when the ATR core was briefly at low power, so AGR-2 irradiation data are not used for physics and thermal calculations. Also, two cycles, 150A and 153B, are Power Axial Locator Mechanism (PALM) cycles when the ATR power is higher than during normal cycles. During the first PALM cycle, 150A, the experiment was temporarily moved from the B-12 location to the ATR water canal and during the second PALM cycle, 153B, the experiment was temporarily moved from the B-12 location to the I-24 location to avoid being overheated. During the “Outage” cycle, 153A, seven flow meters were installed downstream from seven Fission Product Monitoring System (FPMS) monitors to measure flows from the monitors and these data are included in the NDMAS database.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Corrosion Testing of Refractory inContact with Molten Glasses Designed for Waste Vitrification - VSL Touchpoint Matrix Glasses

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests. Refractory corrosion is generally reported as physical material loss, measured in units of distance (e.g., inch) or as physical material loss rate, measured in units of distance per time (e.g., inch/day). In post-operational melters, the refractory corrosion is measured directly, sometimes reported as corrosion depth. Crucible tests are used in the laboratory to accelerate the refractory corrosion to facilitate a meaningful measurement in a commensurate amount of time. Crucible tests are particularly useful in understanding refractory corrosion across a large glass composition space, where operational testing would be prohibitive. Some of the critical parameters that are known to influence refractory corrosion by molten glass in a crucible test are temperature, system redox, molten salt phases, glass chemistry, and test duration. The majority of data collected for Monofrax® K-3 (hereafter referred to as K-3) corrosion is from crucible tests, but a small amount comes directly from scaled and production melters. Crucible test data has been collected under varying conditions, whereas data collected from operational melters is relatively fewer and represents conditions specific to the melter campaign. The result is that the published data can be grouped and analyzed in multiple ways, not all of which are readily comparable. The Standard Test Method for Isothermal Corrosion Resistance of Refractories to Molten Glass (ASTM C621) outlines the general guidelines used across industry. That method describes a sealed, static test in which the surface area of the refractory coupon and the volume of glass are fixed. A significant portion of the crucible data pertaining to nuclear waste glasses has been collected in a modified configuration; the most notable differences being the surface area of the refractory coupon to volume of the glass and use of a method for bubbling the melt. To our knowledge, the influence of those parameters on the refractory corrosion has not been quantified. In this work, it was determined that static tests and bubbled tests would be performed. Savannah River National Laboratory (SRNL) was tasked with setting up and performing static testing while PNNL was tasked with setting up and performing bubbled testing. Initial activities were performed to establish laboratory methods that reproduce data comparable to existing data sets of K-3 refractory corrosion by low activity waste (LAW) and high-level waste (HLW) glass compositions. Later activities were focused on refining the test parameters to establish a standard test practice to be used between Laboratories and collecting additional data to be used in the enhanced waste glass model development. This document serves primarily to convey the refractory loss measurement results from corrosion testing of K-3 refractory with waste glass compositions developed for use in the WTP melters. The data will be used in the enhanced property/composition models being developed for waste glass vitrification and melter operations.

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

Brine Transport Experiments in Granular Salt

To gain confidence in the predictive capability of numerical models, experimental validation must be performed to ensure that parameters and processes are correctly simulated. The laboratory investigations presented herein aim to address knowledge gaps for heat-generating nuclear waste (HGNW) disposal in bedded salt that remain after examination of prior field and laboratory test data. Primarily, we are interested in better constraining the thermal, hydrological, and physicochemical behavior of brine, water vapor, and salt when moist salt is heated. The target of this work is to use run-of-mine (RoM) salt; however during FY2015 progress was made using high-purity, granular sodium chloride.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Reproducible Data Processing Research for the CABRI R.I.A. experiments Acoustic Emission signal analysis

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission (CEA) designed to study the behaviour of fuel rods at high burnup under Reactivity Initiated Accident (R.I.A.) conditions such as the scenario of a control rod ejection. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream from the test device. Studies carried out on the last fourteen tests showed the interest in carrying out temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena. We want presently to return to this rich data in order to have a new point of view by applying modern signal processing methods. Such an antecedent works resumption leads to some difficulties. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not clear in reproducing from the former studies since the people who were in charge of the original work have left the laboratory and it is not easy when time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup. Thus we decided to consolidate the availability of the data and its manipulation in order to provide a robust data processing workflow to the experimentalists before doing any further investigations. To tackle this issue of strong links between data, treatments and the generation of documents, we adopted a Reproducible Research paradigm. We shall first present the tools chosen in our laboratory to implement this workflow and, then we shall describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the last fourteen CABRI R.I.A. tests. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Preliminary Results of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process

Idaho National Laboratory is building a test bed to allow researchers the opportunity to study nuclear fuel processing operations. This includes studying the solvent extraction process and the use of centrifugal contactors. The goal of this project is to develop a system that utilizes non-traditional measurement sources such as vibration, acoustics, current, color, flow, and temperature in conjunction with data-based, machine learning techniques that will allow for signal discovery. This multi-sensor data supports the development of safeguards by design, provides operator process awareness, and aids in the discovery of process anomalies. This paper highlights some of the preliminary results from initial data collection campaigns and shares some of the lessons learned.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Jet tomography of hot and cold nuclear matter [Slides]

Jets are probes of microscopic structures of nuclear matter. With inclusive jets and hadrons, the transport approach is used to determine the QGP jet transport parameter. Much more information contained in k T -dependent observables, require to go beyond qˆ approximation and new tools. Test in-medium jet theory with existing eA data and future EIC.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Inverse Sensitivity/Uncertainty Methods Development for Nuclear Fuel Cycle Applications

The Standardized Computer Analyses for Licensing Evaluation (SCALE) software package developed at the Oak Ridge National Laboratory includes codes that propagate uncertainties available in the nuclear data libraries to compute uncertainties in nuclear application performance parameters (responses). We report on our recently commenced eorts to extend this capability to develop an inverse sen- sitivity/uncertainty (IS/U) methodology that identies the improvements in nuclear data that are needed to compute application responses within prescribed tolerances, while minimizing the cost of such data improvements. We report on our progress to date and present a simple test case for our method. Our methodology is directly applicable to thermal and intermediate neutron energy systems because it addresses the implicit neutron resonance self-shielding eects that are essential to accurate modeling of thermal and intermediate systems. This methodology is likely to increase the eciency of nuclear data eorts.

Arbanas, Goran↗
DOE OSTI2011

AGC-1 Pre-Irradiation Data Report Status

The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All samples in the experiment will be fully characterized before irradiation, irradiated in the Advanced Test Reactor (ATR), and then re-examined to determine the irradiation induced changes to key materials properties in the different graphite grades. The information generated during the AGC experiment will be utilized for NRC licensing of NGNP reactor designs, shared with international collaborators in the Generation IV Information Forum (GIF), and eventually utilized in ASME design code for graphite nuclear applications. This status report will describe the process the NGNP Graphite R&D program has developed to record the AGC1 pre-irradiation examination data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Network sensitivity solutions for regional moment-tensor inversions

Well-resolved moment-tensor solutions reveal information about the sources of seismic waves. In this paper,we introduce a newly of assessing confidence in the regional full moment-tensor inversion via the introduction of the network sensitivity solution (NSS). The NSS takes into account the unique station distribution, frequency band, and signal-to-noise ratio of a given event scenario. The NSS compares both a hypothetical pure source (for example, an explosion or an earthquake) and the actual data with several thousand sets of synthetic data from a uniform distribution of all possible sources. The comparison with a hypothetical pure source provides the theoretically best-constrained source-type distribution for a given set of stations; and with it, one can determine whether further analysis with the data is warranted. The NSS that employs the actual data gives a direct comparison of all other source types with the best fit source. In this way, one can choose a threshold level of fit in which the solution is comfortably constrained. The method is tested for the well-recorded nuclear test, JUNCTION, at the Nevada Test Site. Sources that fit comparably well to a hypothetical pure explosion recorded with no noise at the JUNCTION data stations have a large volumetric component and are not described well by a double-couple (DC) source. The NSS using the real data from JUNCTION is even more tightly constrained to an explosion because the data contain some energy that precludes fitting with any type of deviator source. We also calculate the NSS for the October 2006 North Korea test and a nearby earthquake, where the station coverage is poor and the event magnitude is small. As a result, the earthquake solution is very well fit by a DC source, and the best-fit solution to the nuclear test (M w 4.1) is dominantly explosion.

58 GEOSCIENCES↗
DOE OSTI2017

CURRENT STATUS OF THE CHARACTERIZATION OF RPV MATERIALS HARVESTED FROM THE DECOMMISSIONED ZION UNIT 1 NUCLEAR POWER PLANT

The decommissioning of Units 1 and 2 of the Zion Nuclear Power Station in Zion, Illinois, after ~ 15 effective full-power years of service presents a unique opportunity to characterize the degradation of in-service reactor pressure vessel (RPV) materials and to assess currently available models for predicting radiation embrittlement of RPV steels [1-3]. Moreover, through-wall thickness attenuation and property distributions are being obtained and the results to be compared with surveillance specimen test data. It is anticipated that these efforts will provide a better understanding of materials degradation associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service and subsequent license renewal. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the U.S. Department of Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated procurement of materials, components, and other items of interest from the decommissioned Zion NPPs. In this report, harvesting, cutting sample blocks, machining test specimens, test plans, and the current status of materials characterization of the RPV from the decommissioned Zion NPP Unit 1 will be discussed. The primary foci are the circumferential, Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B base metal from the intermediate shell harvested from a region of peak fluence (0.7 X 1019 n/cm2, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following the determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing are being performed to characterize the through-thickness mechanical properties of base metal and beltline-weld materials. In addition to mechanical properties, microstructural characterizations are being performed using various microstructural techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy

Rosseel, T. M.↗