Idaho National Laboratory Research Articles

Geovisualization applications to examine and explore high-density and hierarchical critical infrastructure data

Abstract. The geospatial research and development team in the National and Homeland Security Division at Idaho National Laboratory was tasked with providing tools to derive insight from the substantial amount of data currently available – and continuously being produced – associated with the critical infrastructure of the US. This effort is in support of the Department of Homeland Security, whose mission includes the protection of this infrastructure and the enhancement of its resilience to hazards, both natural and human. We present geovisual-analytics-based approaches for analysis of vulnerabilities and resilience of critical infrastructure, designed so that decision makers, analysts, and infrastructure owners and managers can manage risk, prepare for hazards, and direct resources before and after an incident that might result in an interruption in service. Our designs are based on iterative discussions with DHS leadership and analysts, who in turn will use these tools to explore and communicate data in partnership with utility providers, law enforcement, and emergency response and recovery organizations, among others. In most cases these partners desire summaries of large amounts of data, but increasingly, our users seek the additional capability of focusing on, for example, a specific infrastructure sector, a particular geographic region, or time period, or of examining data in a variety of generalization or aggregation levels. These needs align well with tenets of in-formation-visualization design; in this paper, selected applications among those that we have designed are described and positioned within geovisualization, geovisual analytical, and information visualization frameworks.

On the flow induced vibration of an externally excited nuclear reactor experiment

An in-pile, drop-in experiment design is presently being designed and studied for the near-term deployment within the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL); this experiment is termed the Miniplate-1 Large-B (MP-1 LB) Experiment. A number of explicit studies are performed during the design- and safety-related stage. Traditionally, a clear and logical methodology has been developed and utilized for analyses such as hydraulics, thermal-loads, mechanical loads, and others for these experiments. Recently a small component from a different experiment assembly mechanically separated while in the reactor’s core. While this experiment didn’t compromise the safety of the reactor, it led to a higher-level question which centered on whether the appropriate level of consideration was being made toward the fluid–structure-interactions of these experiments. The outcome yielded separate flow test experiments of like-for-like geometry in an experimental loop located at Oregon State University which produces experimental data compliant with applicable parts and requirements to ASME’s NQA-1 2008, 2009a standard – suitable for benchmark evaluation. The objectives of this study are to (1) develop a process for handling and interpreting the mechanical response of the test elements during hydraulic testing, to (2) characterize the motion of a specific test element during a flow test which imposes a wide range of hydraulic conditions, and (3) provide objective observations toward the potential safety related implications that are tied to the synthesized data. The outcome of this study has led to a confident process in inspecting the experimental data, synthesized it for interpretation, identified several unique hydraulic characteristics of the experiment design which were previously unknown, and demonstrated that the likelihood for mechanical failure resulting from fluid-structure-interactions in the reactor is far below any criterion for concern of the element’s safety.

Recent accomplishments of the fusion safety program at the Idaho National Laboratory

The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences’ (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL.

(Invited) Hydrogen Is an AWSM Energy Materials Network!

The Energy Materials Network (EMN) is a Department of Energy (DOE) network of national lab-led consortium aimed at accelerating the development and commercial deployment of novel materials by enhancing the accessibility of unique material research resources at the national laboratories to external stakeholders, such as academia and industry. The HydroGEN EMN was launched in October 2016 by the DOE Office of Energy Efficiency and Renewable Energy (EERE) Fuel Cell Technologies Office (FCTO). HydroGEN (https://www.h2awsm.org/) EMN is a six-lab consortium, led by the National Renewable Energy Laboratory (NREL). It currently comprises six core national laboratories: NREL, Sandia National Laboratory (SNL), Lawrence Berkeley National Laboratory (LBNL), Idaho National Laboratory (INL), Lawrence Livermore National Laboratory (LLNL), and Savannah River National Laboratory (SRNL). HydroGEN aims to accelerate the discovery and development advanced water splitting materials (AWSM) for sustainable, large-scale hydrogen production, in order to more effectively enable the widespread commercialization of hydrogen and fuel cell technologies. With the rollouts of fuel cell electric vehicles (FCEVs) by major automotive manufacturers underway, enabling AWS technologies for the widespread production of affordable, sustainable hydrogen becomes increasingly important. Hydrogen is a unique energy carrier in that it can be produced from a number of diverse pathways, utilizing a variety of domestically available feedstocks, including natural gas, biomass, and water. Advanced water splitting (AWS) technologies, including advanced electrolysis (low and high temperature), photoelectrochecmical (PEC) and solar thermochemical (STCH) routes, are some of the more versatile pathways, and will play a significant role in long-term, high volume sustainable production. The HydroGEN Consortium offers an extensive collection of materials research capabilities for addressing RD&D challenges in efficiency, durability and cost. Leveraging the HydroGEN Consortium’s staff of leading technical experts and broad collection of resource capabilities is expected to advance the maturity and technology readiness levels in all the advanced water splitting technologies. In June 2017, DOE EERE FCTO announced the award of 18 new HydroGEN seedling projects, and one project focused on benchmarking advanced water splitting technologies. These 19 new projects utilized 44 unique capabilities across the six HydroGEN core labs. Furthermore, it is a nationwide R&D effort that comprises 10 national labs, 6 companies, and 22 universities. This presentation will provide an overview of the HydroGEN EMN consortium and highlight some of the advanced water electrolysis projects that are focused on hydrogen and oxygen evolution catalysis.

Electrochemistry of Iodide in LiCl-KCl Molten Salts and Anionic Chemla Effect: An Overview

Several radioactive iodine isotopes are formed as fission products in the nuclear fuel, and retained within the fuel matrix by the fuel cladding as a containment. During reprocessing of the used fuels by the pyroprocessing route, a significant fraction of iodine was reported to be retained in the molten salt as iodide (I-). High concentration of iodine leads to significant deterioration of the Pt anode during electrolytic reduction of the oxide spent fuels. [[i]] The electrochemistry of iodine/iodide couple has been extensively studied in aqueous and non-aqueous conditions because of its application in dye-sensitized solar cells, and synthetic chemistry. The physical and electrical properties of the molten salt systems depend on the structure and interactions of the constituents. The structural arrangements that are present in molten salts can be viewed as intermediates between discrete chemical bonds and periodic crystalline lattices[[ii]]. The inter-atomic interactions determine the local ordering of the molten salt. The electrochemical properties are significantly influenced by the structural characteristics of the molten salt. The ion size and type in a molten salt system such as KX-LiX would affect the electrochemical stability. The ordering and structure of the melt could be described based on the entropy. The entropy (ΔS) of the electrolysis of melt can be determined using the relation: ΔS = nF(∂E/∂T)P (1) Where, n = number of electrons, (∂E/∂T)P = change in electrochemical window with temperature at constant pressure. Low entropy values indicate higher order and enhanced attractive interaction of the species in the molten salt. Generally, an increase in the entropy was observed with increase in the anion size of unary molten salts[[iii]]. The electrolysis potential turns out to be more anodic as the anion size increases due to weaker Coulombic interaction between the anion-cation pairs as well as increased repulsion between the larger anions[[iv]]. When two types of cations are present in molten salt with different sizes and different charge densities, an asymmetric polarization of anions is anticipated which may result in the electrostatic stability of the mixture. When two different types of anions such as chloride and iodide are present along with different cations, the asymmetric polarizations of the ions and their effect on the physical and electrochemical properties are not well documented. The Chemla effect[v] refers to a phenomenon where the internal mobility of a larger cation is higher than that of a smaller cation at high concentrations of the larger cation in a binary salt mixture such as LiCl-KCl that consists of large and small cations with a common anion. This effect could lead to compositional gradient of ions in an electrochemical cell where a local shift in the composition from liquidus range may cause solid precipitation of salt[[vi]]. Chemla effect has been reported for ternary cation systems such as (Na,K,Cs)Cl [[vii]] and also for anions in Li(Cl,NO3)[[viii]]. Presence of iodide at above a threshold concentration could result in anionic Chemla effect where the internal mobility of iodide could be higher than that of chloride. The other possible scenario could be at low concentration of chloride, the mobility of chloride could be larger than that of iodide, as observed in the Li(Cl,NO3). However, in normal pyroprocessing condition, the chloride concentration will not be lower than that of iodide. The redox reactions of iodide are given as: 2I- ↔ I2 + 2e- (2) 3I- ↔ I3 - + 2e- (3) With the chloride addition, the redox reactions are given as: I- + 2Cl- ↔ [ICl2]- + 2e- (4) I3 - + 6Cl- ↔ 3[ICl2]- + 4e- (5) This presentation will give an overview of the electrochemistry of iodide in the LiCl-KCl molten salt system that is relevant to the pyroprocessing of used nuclear fuels based on the existing database from the published literature, and new preliminary results obtained at the University of Idaho in collaboration with the Idaho National Laboratory. [i] S. M. Frank, P.K. Tripathy, S.D. Herrman, Global 2013, Salt Lake City, Sep.29 – Oct. 3, 2013 [ii] J.D. Martin, S.J. Goettler, N. Fosse, L. Iton, Nature, 419 (2002) 381 [iii] M. Chemla, I. Okada, Electrochimica Acta 35 (1990) 1761 [iv] I.K. Delimarskii, B.F. Markov, Electrochemistry of Fused Salts, The Sigma Press Publishing, Washington, DC, 1961 [v] J. Pdrie' and M. Chemla, C. R. Acad. Sci. 250, 3986 (1960). [vi] R. Takagi, H. Shimotake, K. J. Jensen, J. Electrochem. Soc., 131 (1984) 1280 [vii] M. Matsumiya, H. Matsuura, R.Takagi, Y. Okamoto, Journal of The Electrochemical Society, 147 (11) 4206-4211 (2000) [viii] A. Endoh, I.Okada, J. Electrochem. Soc., Vol. 137, No. 3, March 1990

Investigations of the Structural and Electrochemical Properties of Overcharged Li-Ion Batteries

Traditional monitoring of electrochemical cells and batteries has been limited to observation of voltage and temperature of the cell or cells. While this monitoring can be very robust, there are limits to how predictive voltage and temperature behavior can be prior to a thermal runaway event. Ultimately, voltage and temperature changes are often lagging symptoms of battery failure, and by the time a noticeable change is detected it is too late to arrest cell failure with intervention or maintenance. Further, instantaneous voltage and temperature monitoring are often inadequate to determine the state of a battery at rest, particularly if the battery has been previously subjected to an abusive condition. Knowledge of the level of stability of a damaged battery would allow for both safer and more efficient handling of the abused battery. This work examines the application of Electrochemical Impedance Spectroscopy (EIS) and Differential Capacity calculations (dQ/dV) as tools for determining the state of stability (SOS) of an electrochemical cell or battery. The cells used for this study were commercial 10 Ah NMC cells and 10Ah LFP cells subjected to electrical abuse coupled with EIS monitoring. This aims to not only provide a deeper understanding of how abused cells and batteries fail, but also form the technical basis of a tool that could ultimately be used to interrogate cells of an unknown stability and even monitor active cells for early signs of damage or failure. Fast impedance monitoring hardware previously developed at Idaho National Laboratory is used to provide not only monitoring after an abusive battery test but also look for changes in the cell while abusive conditions are applied. Differential capacity calculations are explored both before tests and after moderate levels of overcharge to explore any observable changes that may be monitored during charge and discharge operations. Lastly, X-ray diffraction measurements were performed on the cathode/anode before and after the abuse test to understand the effect of overcharge in the structural changes. Figure 1 presents the measurements for the NMC cells that were partially abused from 20-60% OC. These three markers could then be potentially applied to cross-examine batteries of an unknown stability as well as provide the basis for an active diagnostic method as part of a battery management system. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525. Figure 1

Application of regression, variance, and density based global sensitivity methods to integrated VERA-CS and BISON simulations

Uncertainty quantification (UQ) and sensitivity analyses (SA) are performed for coupled simulations between VERA-CS, a coupled pin resolved neutron transport and subchannel thermal hydraulics code, and the fuel performance code BISON. The interface between VERA-CS and BISON is performed in a multiphysics environment known as the LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Laboratory (INL). A focus is placed on using a variety of SA measures, including two regression based (Pearson and Spearman), one variance based (Sobol indices), and three moment independent measures (Delta moment independent measures with L1, L2, and L∞ norms). The problem under inspection is a single assembly depletion case for three fuel cycles. The figures of merit are the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). SA results show MDNBR to be linear with consistent rankings throughout the fuel cycles. MFCT is linear, but with a change in rankings at the switch from open gap to closed gap models. GCPP is nonlinear at intermediate states that coincide with the onset of contact between fuel and cladding. These nonlinear states allow for the showcasing of higher order SA measures over first order methods.

Overview of the incompressible Navier–Stokes simulation capabilities in the MOOSE framework

The Multiphysics Object Oriented Simulation Environment (MOOSE) framework is a high-performance, open source, C++ finite element toolkit developed at Idaho National Laboratory. MOOSE was created with the aim of assisting domain scientists and engineers in creating customizable, high-quality tools for multiphysics simulations. While the core MOOSE framework itself does not contain code for simulating any particular physical application, it is distributed with a number of physics “modules” which are tailored to solving e.g. heat conduction, phase field, and solid/fluid mechanics problems. In this report, we describe the basic equations, finite element formulations, software implementation, and regression/verification tests currently available in MOOSE’s navier_stokes module for solving the Incompressible Navier-Stokes (INS) equations.

Predicting Adversarial Cyber-Intrusion Stages Using Autoregressive Neural Networks

Current cybersecurity approaches are response-driven and ineffective, as they do not account for dynamic adversarial movement. Using empirical evidence from observations of two Red Team-Blue Team cybersecurity training exercises held at Idaho National Laboratory and the Michigan Cyber Range, we used four different models to make temporal predictions of how adversaries progress through cyberattacks: nonlinear autoregressive (NAR) neural network, NAR neural network with exogenous input (NARX), NAR neural network for multi-steps-ahead prediction, and autoregressive integrated moving average (ARIMA). The obtained results demonstrate that the trained models can capture different variations in adversarial movement across the two datasets with reliable accuracy.

Tritium permeability measurement in hydrogen-tritium system

Understanding of thermodynamic equilibria of multi-components hydrogen isotopes is required to accurately measure tritium permeability at the expected low tritium partial pressure and non-negligible high hydrogen partial pressure in a fusion blanket system. A gas-driven tritium permeation system that is capable of independently controlling hydrogen and tritium partial pressures was developed at Idaho National Laboratory to accurately measure low partial pressure tritium permeability. The thermodynamic equilibria for hydrogen (H) – tritium (T) permeation through metal are discussed to accurately measure tritium permeability, and the experimental conditions required for evaluating tritium permeability in H-T system are presented.

The effects of radiation chemistry on radiochemistry: when unpaired electrons defy great expectations

In liquid/liquid extraction fuel cycle scenarios, successful metal partitioning depends upon maintaining predictable ligand concentrations and metal valence states in a non-equilibrium system perturbed by the constant input of energy. Energetic particles from nuclide decay interact with solution to yield ions and radicals that react with solutes. These reactions have the potential to interfere with the expected separations chemistry. This paper describes example reactions of the main transient products of aqueous and organic diluents with ligands or metal ions. These reactions tend to decrease ligand concentrations, generate unwanted products, and change metal valence states. It discusses the effects of these reactions on the desired separations, using examples mainly from work conducted at the Idaho National Laboratory Center for Radiation Chemistry Research (CR2).

A Geant4 evaluation of the Hornyak button and two candidate detectors for the TREAT hodoscope

The performance of traditional Hornyak buttons and two proposed variants for fast-neutron hodoscope applications was evaluated using Geant4. The Hornyak button is a ZnS(Ag)-based device previously deployed at the Idaho National Laboratory’s TRansient REActor Test Facility (better known as TREAT) for monitoring fast neutrons emitted during pulsing of fissile fuel samples. Past use of these devices relied on pulse-shape discrimination to reduce the significant levels of background Cherenkov radiation. Proposed are two simple designs that reduce the overall light guide mass (here, polymethyl methacrylate or PMMA), employ silicon photomultipliers (SiPMs), and can be operated using pulse-height discrimination alone to eliminate background noise to acceptable levels. Geant4 was first used to model a traditional Hornyak button, and for assumed, hodoscope-like conditions, an intrinsic efficiency of 0.35% for mono-directional fission neutrons was predicted. The predicted efficiency is in reasonably good agreement with experimental data from the literature and, hence, served to validate the physics models and approximations employed. Geant4 models were then developed to optimize the materials and geometries of two alternatives to the Hornyak button, one based on a homogeneous mixture of ZnS(Ag) and PMMA, and one based on alternating layers of ZnS(Ag) and PMMA oriented perpendicular to the incident neutron beam. For the same radiation environment, optimized, 5-cm long (along the beam path) devices of the homogeneous and layered designs were predicted to have efficiencies of approximately 1.3% and 3.3%, respectively. For longer devices, i.e., lengths larger than 25 cm, these efficiencies were shown to peak at approximately 2.2% and 5.9%, respectively. Moreover, both designs were shown to discriminate Cherenkov noise intrinsically by using an appropriate pulse-height discriminator level, i.e., pulse-shape discrimination is not needed for these devices.

Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization

The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage of analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.

Automated Work Packages: Capabilities of the Future

As part of the ongoing efforts at the U.S. Department of Energy’s Light Water Reactor Sustainability Program, Idaho National Laboratory is conducting several research projects in collaboration with the nuclear industry to improve the reliability, safety, and economics of the nuclear power industry, especially as the nuclear power plants extend their operating licenses to 80 years. One of these projects is the automated work package (AWP) project. An AWP is an electronic intelligent and interactive work package. It uses plant condition, resources status, and user progress to adaptively drive the work process in a manner that increases efficiency while reducing human error. To achieve this mission, the AWP acquires information from various systems of a nuclear power plant and incorporates several advanced instrumentation and control technologies along with modern human factors techniques.With the current rapid technological advancement, it is possible to envision several available or soon-to-be-available capabilities that can play a significant role in improving the work package process. As a pilot project, the AWP project develops a prototype of an expanding set of capabilities and evaluates them in an industrial environment. While some of the proposed capabilities are based on using technological advances in other applications, others are conceptual; thus, they require significant research and development to be applicable in an AWP. The scope of this paper is to introduce a set of envisioned capabilities, their need for the industry, and the industry difficulties they resolve.

Testing fast reactor fuels in a thermal reactor

The present lack of a domestic fast neutron flux irradiation capability combined with continued development of fast reactor fuels in the U.S. motivated an innovative engineering solution to utilize a unique neutron flux tailoring capability in the Advanced Test Reactor at the Idaho National Laboratory. To achieve the objectives of the fast reactor fuel irradiation tests, the incident neutron flux was hardened substantially by placing fueled irradiation capsules inside specially designed cadmium shrouds. Use of cadmium prevents thermal neutrons from reaching the fuels being tested and alleviates the plutonium self-shielding that would normally arise during irradiations of high density, highly enriched fuels in a thermal neutron spectrum.The present paper illustrates the profound effect this engineered solution has on the efficacy of the experiments. Based on the comparison of post-irradiation measurements of the columnar grain region in fast reactor mixed-oxide fuels with fuel performance calculations, it is demonstrated that thermal conditions achieved in these cadmium-shrouded fuel experiments are substantially prototypic of a sodium fast reactor and are suitable for concept-screening tests supporting development of new fast reactor fuels. It is also shown that if the experiments were conducted in an unmodified ATR neutron spectrum, gross plutonium self-shielding would cause a strong depression of the fission power at the fuel centerline preventing fuel restructuring, a hallmark feature of mixed oxide fuel behavior under fast reactor conditions.Recognizing the need for testing metallic fuels for fast reactors, the impact of neutron flux energy spectrum on the radial temperature distribution in metallic fuel is investigated. It is shown that the use of Cd shrouds allows to attain radial temperature distributions nearly identical to those that exist in an SFR.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

The AGR-3/4 fission product transport irradiation experiment

AGR-3/4 was the combined third and fourth planned irradiations for the U.S. Department of Energy (DOE) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The primary purpose of the AGR program is to support the development and qualification of tristructural isotropic (TRISO)-coated particle fuel for use in High Temperature Gas-cooled Reactors. AGR-3/4 was designed as a fission product transport irradiation experiment whose specific objectives were to: (1) irradiate fuel containing UCO (uranium oxycarbide) designed-to-fail (DTF) fuel particles that provide a fixed source of fission products for subsequent transport through compact matrix and structural graphite materials; (2) assess the effects of sweep gas impurities on fuel performance and fission product transport; (3) provide irradiated fuel and material samples for post-irradiation examination (PIE) and post-irradiation heating; and (4) support the refinement of fuel performance and fission product transport models. The AGR-3/4 test train was irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for 369.1 effective full power days from December 2011 to April 2014. The experiment was successful in achieving its specification goals in terms of burnup and fast fluence levels reached at the end of irradiation and fuel temperature levels maintained throughout irradiation: peak compact burnup reached 15.27% fissions per initial heavy-metal atom and peak compact fast fluence reached 5.32 × 1025 n/m2 (E > 0.18 MeV), while the time-average volume-average temperatures of the compacts ranged from 854 to 1345 °C. Fission product release–to–birth ratios reached values in the 10−4–10−3 range early during irradiation as the DTF particles started to fail. Subsequent post-irradiation examination will provide information on fission product distributions in matrix and core graphite materials, enabling refinement of fission product transport models.

The IAEA CRP on HTGR uncertainties: Sensitivity study of PHISICS/RELAP5-3D MHTGR-350 core calculations using various SCALE/NEWT cross-section sets for Ex. II-1a

Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. A Coordinated Research Program (CRP) supervised by the International Atomic Energy Agency was started to investigate the various uncertainty quantification methodologies for High Temperature Gas-cooled Reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Best-estimate results for the prismatic single block (Ex. I-2a) and super cell models (Ex. I-2c) were obtained using the SCALE 6.2.0 two-dimensional lattice code NEWT. A reference spectrum was obtained with Serpent 2.1.27 for the single block, super cell and core models. The flux spectrum in the system of interest plays a primary role in the quantification of uncertainties caused by cross sections. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex. I-2a and various models of Ex.I-2c are utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the Idaho National Laboratory (INL) coupled code PHISICS/RELAP5-3D. It is observed that axial shape of the core power density does not vary significantly with the various lattice cell libraries utilized. The use of cross section libraries originating from super cells induces changes of the core power density by 1–10% radially as compared to the Ex.I-2a cross sections. The magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models. A lattice flux spectrum resembling the core spectrum is hence necessary for correct predictions in nodal core calculations.

Optimization of heat treatment and calibration procedures for high temperature irradiation resistant thermocouples

ABSTRACTA detailed description is provided for the optimization of heat treatment and calibrating procedure for Idaho National Laboratory’s high temperature irradiation resistant thermocouples (HTIR-TC). Also discussed is the implication of the procedures on the overall performance of the HTIR-TC finished product; in particular, for the case of long lead sections that are typical of sensors deployed in nuclear reactors. The effect on the measurement accuracy of fluctuations of the reference temperature and localized heating along the sensor lead is also investigated. A calibration graphical user interface (GUI) and accompanying script is presented for unifying the calibration process. A fifth-order polynomial function was found to be best for fitting HTIR-TC calibration data. The performance of the HTIR-TCs subject to the optimized heat treatment and calibration procedures is shown to be robust and consistent with other TC industry standards. Electromotive force curves versus length of the TC are presented to clarify positioning of TCs in general and how its effect on the calibrated temperatures.

Full‐Scale Drinking Water System Decontamination at the Water Security Test Bed

The EPA's Water Security Test Bed (WSTB) facility is a full-scale representation of a drinking water distribution system. In collaboration with the Idaho National Laboratory (INL), EPA designed the WSTB facility to support full-scale evaluations of water infrastructure decontamination, real-time sensors, mobile water treatment systems, and decontamination of premise plumbing and appliances. The EPA research focused on decontamination of 1) Bacillus globigii (BG) spores, a non-pathogenic surrogate for Bacillus anthracis and 2) Bakken crude oil. Flushing and chlorination effectively removed most BG spores from the bulk water but BG spores still remained on the pipe wall coupons. Soluble oil components of Bakken crude oil were removed by flushing although oil components persisted in the dishwasher and refrigerator water dispenser. Using this full-scale distribution system allows EPA to 1) test contaminants without any human health or ecological risk and 2) inform water systems on effective methodologies responding to possible contamination incidents.