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DOE OSTI2015

Preliminary Design of Critical Function Monitoring System of PGSFR

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation control and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Dynamic Braking System of a Tidal Generator

Renewable energy generation has experienced significant cost reductions during the past decades, and it has become more accepted by the global population. In the beginning, wind generation dominated the development and deployment of renewable energy; however, during recent decades, photovoltaic (PV) generation has grown at a very significant pace due to the tremendous decrease in the cost of PV modules. The focus on renewable energy generation has now expanded to include new types with promising future applications, such as river and tidal generation. The input water flow to these types of resources is more predictable than wind or solar generation. The data used in this paper is representative of a typical river or tidal generator. The analysis is based on a generator with a power rating of 40 kW. The tidal generator under consideration is driven by two sets of helical turbines connected to each side of the generator located in between the turbines. The generator is operated in variable speed, and it is controlled to maximize the energy harvested as well as the operation of the turbine generator. The electrical system consists of a three-phase permanent magnet generator connected to a three-phase passive rectifier. The output of the rectifier is connected to a DC-DC converter to match the rectifier output to the DC bus voltage of the DC-AC inverter. The three-phase inverter is connected to the grid, and it is controlled to provide a good interface with the grid. One important aspect of river and tidal generation is the braking mechanism. In a tidal generator, the braking mechanism is important to avoid a runaway condition in case the connection to the grid is lost when there is a fault in the lines. A runaway condition may lead to an overspeed condition and cause extreme stresses on the turbine blade structure and eventual disintegration of the mechanical structure. In this paper, the concept of the dynamic braking system is developed and investigated for normal and abnormal operations. The main objective is to optimize the performance under emergency braking while designing the system to be as simple as possible to avoid overdesigning the power electronics or exceeding the target budget.

13 HYDRO ENERGY↗

DOE OSTI2010

Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the realizable k-epsilon model with two-layer all y+ wall treatment performs better than the other k-epsilon and k-omega turbulence models when compared to the experimental results and the Reynolds stress transport turbulence model results. A scaling analysis was developed to address the distortions introduced by the CFD model in simulating the physical phenomena inside the RCCS system with respect to the full plant configuration. The scaling analysis demonstrated that both the experimental facility and the CFD model achieve a satisfactory resemblance of the main flow characteristics inside the RCCS cavity region, and convection and radiation heat exchange phenomena are properly scaled from the actual plant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Thermal-Hydraulic and Neutronic Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

In this paper, we discuss liquid-fuel molten salt (cooled) reactors (MSRs); how they will operate under normal, transient, and accident conditions; and the results of an expert elicitation to determine the corresponding thermal-hydraulic and neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy, and industry to develop or identify modeling functionalities and tools required to carry out confirmatory and licensing analyses that examine the validity and accuracy of an applicant's calculations and help determine the margin of safety in plant design. The NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel comprising a group of subject matter experts met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures of merit (FoMs) (e.g., salt temperature, velocity, and composition). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs, with an emphasis on the thermal-hydraulic takeaways. It was concluded that compared to light water reactors, the lack of high-pressure operation, energetic break flow, depressurization, and quench front tracking may simplify some aspects of an MSR analysis. However, MSRs have new phenomena both for a license applicant and NRC confirmatory analysis. There is a need for enhanced understanding of physical properties for MSRs that encompass several individual thermophysical properties, including thermal conductivity, viscosity, specific heat, density, optical properties, thermodynamic properties, volatilities, solubilities, etc. Salt composition is closely linked to both these properties and the neutronics of the system. Additionally, the large number of MSR concepts and system designs means that there is wide variation in the potential modeling needs for these systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Grizzly: A Multi-scale and Multi-Physics Tool to Model Aging of Nuclear Power Plant Components

Grizzly is a multi-scale multi-physics tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy's Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL's MOOSE framework, which enables multi-physics simulations in a massively parallel computing environment. The reactor pressure vessel (RPV) has been chosen as the initial application for Grizzly. The primary concern for the RPV is that it becomes increasingly susceptible to fracture initiation at the sites of pre-existing flaws over time. This is because exposure to irradiation and high temperatures embrittles the steel. Assessing the safety of an RPV under transient conditions involves simulations at multiple scales. At the engineering scale, strongly coupled equations of heat conduction and solid mechanics are solved to simulate the global response of the RPV to accident conditions. Sub-models are used to represent regions with pre-existing flaws and the stress intensity factors are evaluated from the domain integrals. Crack propagation is then determined using a model that provides the temperature and neutron fluence dependent fracture toughness of the RPV steel. The ability of Grizzly to provide accurate engineering scale estimates of RPV safety beyond 60 years of operation strongly relies on the predictability of the embrittlement model. Multiple models for embrittlement have been proposed, and the physically based empirical model is implemented in Grizzly. However, the formation of late blooming phases greatly reduces the confidence on the extrapolated values obtained from embrittlement model, which is further exacerbated by the absence of real field data for validation. Experimentally validated lower length scale models can significantly improve the predictability of the embrittlement model beyond the calibrated range and the development of such models is another ongoing effort in Grizzly. The modeling activity in Grizzly spans across different length scales, and can be divided into: (i) Irradiation induced microstructure evolution at the atomistic and single crystal length scale using molecular dynamics, atomistic kinetic Monte-Carlo, phase-field and rate theory; (ii) Effect of the irradiation induced defects on the flow stress and fracture processes at the polycrystalline scale using crystal plasticity. Subsequently, the prediction from these models can be used to obtain irradiation induced ductile-to-brittle transition temperature (DBTT) shifts from specimen-scale simulations; (iii) Global thermo-mechanical response of the RPV under normal and accident scenarios, and probabilistic risk analyses at the engineering scale using the DBTT shifts obtained from the lower length scale models. The item (i) has been discussed in a previous report. In this extended abstract an overview of items (ii) and (iii) is provided. The extended abstract summarizes the current status of the modeling effort under Grizzly. At the engineering scale, reduced order model development is underway using the statistical tool RAVEN. This will allow probabilistic analysis considering a wide range of flaw topologies in a computationally efficient manner. Coupling the irradiation induced defect growth predictions to crystal plasticity is also in progress. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

First on-line detection of radioactive fission isotopes produced by laser-accelerated protons

Abstract The on-going developments in laser acceleration of protons and light ions, as well as the production of strong bursts of neutrons and multi- $$\hbox {MeV}$$ MeV photons by secondary processes now provide a basis for novel high-flux nuclear physics experiments. While the maximum energy of protons resulting from Target Normal Sheath Acceleration is presently still limited to around $$100 \, \hbox {MeV}$$ 100 MeV , the generated proton peak flux within the short laser-accelerated bunches can already today exceed the values achievable at the most advanced conventional accelerators by orders of magnitude. This paper consists of two parts covering the scientific motivation and relevance of such experiments and a first proof-of-principle demonstration. In the presented experiment pulses of $$200 \, \hbox {J}$$ 200 J at $$\approx \, 500 \, \hbox {fs}$$ ≈ 500 fs duration from the PHELIX laser produced more than $$10^{12}$$ 10 12 protons with energies above $$15 \, \hbox {MeV}$$ 15 MeV in a bunch of sub-nanosecond duration. They were used to induce fission in foil targets made of natural uranium. To make use of the nonpareil flux, these targets have to be very close to the laser acceleration source, since the particle density within the bunch is strongly affected by Coulomb explosion and the velocity differences between ions of different energy. The main challenge for nuclear detection with high-purity germanium detectors is given by the strong electromagnetic pulse caused by the laser-matter interaction close to the laser acceleration source. This was mitigated by utilizing fast transport of the fission products by a gas flow to a carbon filter, where the $$\upgamma$$ γ -rays were registered. The identified nuclides include those that have half-lives down to $$39 \, \hbox {s}$$ 39 s . These results demonstrate the capability to produce, extract, and detect short-lived reaction products under the demanding experimental condition imposed by the high-power laser interaction. The approach promotes research towards relevant nuclear astrophysical studies at conditions currently only accessible at nuclear high energy density laser facilities.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2018

Mass, momentum, and energy transfer in supersonic aerosol deposition processes

Aerosol deposition (AD) can lead to the formation of dense coatings via deposition of particles from a gas flow; in AD the aerosol is passed through a converging-diverging nozzle, facilitating inertial particle impaction on a desired substrate at supersonic particle velocities. Unique from thermal spray methods, AD can be applied near room temperature and unique from cold spray, in AD the aerosol is typically at atmospheric pressure upstream of the nozzle. Though AD has been successfully demonstrated previously, a number of aspects related to particle motion in AD systems remain poorly understood. In this work, we simulated compressible flow field profiles and particle trajectories for typical AD working conditions for a slit type converging-diverging nozzle with a planar substrate. In examining the fluid flow profile, we show that the velocity and pressure profiles, as well as the shock structure are sensitive to the upstream and downstream operating pressures of the nozzle. These ultimately affect particle impaction speed. Importantly, in AD, the particle drag regime is dynamic; both particle Knudsen numbers and Mach numbers can vary by orders of magnitude. To aid particle trajectory simulations, we trained a neural network to predict the drag force on the particles based on existing experimental data, theoretical limits, and new direct simulation Monte Carlo (DMSC) results. The neural network based drag law, which depends upon both Mach and Knudsen numbers, shows better agreement with the DSMC simulation data than pre-existing correlations. With it, particle trajectory simulation results reveal that for a given particle density, there exists an optimal particle diameter to maximize particle impaction speed. We also find that in AD particles undergo size dependent inertial focusing, i.e. there is a particular particle diameter where the particle deposition linewidth is minimized. Particles smaller than this diameter are underfocused, and particles larger than this are overfocused, and hence have larger deposition linewidths in both cases. Using trajectory simulations, we additionally develop a framework that can be used to evaluate the position-dependent mass, momentum and kinetic energy fluxes to the deposition substrate for any aerosol size distribution function upstream the nozzle. It is shown for typical aerosol concentrations achievable in the laboratory, the kinetic energy flux can approach a magnitude normally observed in convective heat transfer with phase change, hence translational kinetic energy to thermal energy transfer in AD is likely the key contributor to the formation of dense coatings.

42 ENGINEERING↗

DOE OSTI2016

Pre-Conceptual Design for Northstar ⁹⁹Mo Process Tritium Removal System

In this report we describe a preliminary concept for a Tritium Removal System (TRS) to remove tritium that is generated in the ⁹⁹Mo production process. Preliminary calculations have been performed to evaluate an approximate size for the system. The concept described utilizes well-established detritiation technology based on catalytic oxidation of tritium and tritiated hydrocarbons to water in a high temperature (400 °C) reactor and capture of water in a molecular sieve bed. The TRS concept involves use of a single system that would cycle through each of the seven online target systems and remove tritium that has been accumulated after one week’s run time. The TRS would perform cleanup operations on each target system for a period of approximately 24 hours. This would occur while the system is still online and just prior to target replacement, so tritium levels would at their minimum values for target replacement. In the concept, during normal operation a small fraction (1%) of the helium recirculating in the system would be diverted through the TRS and returned to the flow loop. With this approach sufficient levels of detritiation can be accomplished in a 24 hour period. In the study it was found that because of the need to maintain low oxygen levels in the system (<100 ppm) this increases the size of the catalytic reactor. As a result of this finding, consideration should be given to other methods for removing tritium from the system. Other methods such as catalytic exchange of tritium with an unsaturated organic compound and subsequent trapping on activated carbon or molecular sieve could offer advantages of reducing reactor size and operation at lower reactor temperature. However the most significant advantage of such an approach would be the ability to operate in very low oxygen environments, which would eliminate any concerns for oxidation of the target.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2013

Multiphasic analysis of the temporal development of the distal gut microbiota in patients following ileal pouch anal anastomosis

The indigenous gut microbiota are thought to play a crucial role in the development and maintenance of the abnormal inflammatory responses that are the hallmark of inflammatory bowel disease. Direct tests of the role of the gut microbiome in these disorders are typically limited by the fact that sampling of the microbiota generally occurs once disease has become manifest. This limitation could potentially be circumvented by studying patients who undergo total proctocolectomy with ileal pouch anal anastomosis (IPAA) for the definitive treatment of ulcerative colitis. A subset of patients who undergo IPAA develops an inflammatory condition known as pouchitis, which is thought to mirror the pathogenesis of ulcerative colitis. Following the development of the microbiome of the pouch would allow characterization of the microbial community that predates the development of overt disease. We monitored the development of the pouch microbiota in four patients who underwent IPAA. Mucosal and luminal samples were obtained prior to takedown of the diverting ileostomy and compared to samples obtained 2, 4 and 8 weeks after intestinal continuity had been restored. Through the combined analysis of 16S rRNA-encoding gene amplicons, targeted 16S amplification and microbial cultivation, we observed major changes in structure and function of the pouch microbiota following ileostomy. There is a relative increase in anaerobic microorganisms with the capacity for fermentation of complex carbohydrates, which corresponds to the physical stasis of intestinal contents in the ileal pouch. Compared to the microbiome structure encountered in the colonic mucosa of healthy individuals, the pouch microbial community in three of the four individuals was quite distinct. In the fourth patient, a community that was much like that seen in a healthy colon was established, and this patient also had the most benign clinical course of the four patients, without the development of pouchitis 2 years after IPAA. The microbiota that inhabit the ileal-anal pouch of patients who undergo IPAA for treatment of ulcerative colitis demonstrate significant structural and functional changes related to the restoration of fecal flow. Our preliminary results suggest once the pouch has assumed the physiologic role previously played by the intact colon, the precise structure and function of the pouch microbiome, relative to a normal colonic microbiota, will determine if there is establishment of a stable, healthy mucosal environment or the reinitiation of the pathogenic cascade that results in intestinal inflammation.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2023

Influence of the irradiated pulmonary microenvironment on macrophage and T cell dynamics

The lung is sensitive to radiation, which increases the risk of normal tissue toxicity following radiation therapy. Adverse outcomes include pneumonitis and pulmonary fibrosis, which are thought to result from dysregulated intercellular communication within the pulmonary microenvironment. Macrophage accumulation in the lung following radiation exposure is implicated in these pathogenic outcomes. Here we investigate the role of the microenvironment in macrophage and T cell dynamics. Materials and methods C57Bl/6 mice received 6Gyx5 x-ray irradiation to the right lung. Macrophage and T cell dynamics were investigated in ipsilateral right lungs, contralateral left lungs and compared to those from non-irradiated control subjects at 4-26wk post exposure. Lungs were evaluated by flow cytometry, histology and proteomics. Results In uni-lung irradiated subjects, focal regions of macrophage accumulation were noted in both lungs by 8wk, however fibrotic lesions were observed only in ipsilateral lungs by 26wk. Both infiltrating and alveolar macrophages populations transiently expanded in both lungs, however a transitional CD11b+ alveolar macrophage population persisted only in ipsilateral lungs and expressed lower CD206. Concurrently, arginase-1+ macrophages were noted within areas of macrophage accumulation in ipsilateral but not contralateral lungs at 8 and 26wk post exposure, while CD206+ macrophages were absent from these accumulations. While CD8+ T cells were present in both lungs of irradiated subjects, T regulatory cells were only increased in ipsilateral lungs. Unbiased proteomics analysis of immune cells revealed a substantial number of differentially expressed proteins in ipsilateral lungs when compared to either contralateral or non-irradiated lungs, as well as between contralateral lungs from irradiated subjects and non-irradiated controls. Conclusions Pulmonary macrophage and T cell dynamics are impacted by the microenvironmental conditions that develop in the lung following radiation exposure, both locally and systemically. Our findings demonstrate that while macrophages and T cells infiltrate and expand in both lungs, they diverge phenotypically depending on their environment.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2020

VADER: A Tool for Criticality Safety Validation

The purpose of criticality safety is to prevent any inadvertent criticality from occurring during the handling or storage of fissile material. Calculations are frequently used to demonstrate that a sufficient subcritical margin exists. Validation is a key aspect of the evaluation process, establishing the suitability, accuracy, and associated uncertainty of the computational method and data to be used for the intended application. The validation process is performed by comparing the results of critical experiments with the calculated results from models of the experiments using the computational method to be validated. Laboratory critical experiments are controlled systems that achieve a k eff of approximately 1 in order to investigate the parameters at which such a critical condition is achieved. The validation parameters that are traditionally applied to safety analysis calculations are the bias and the bias uncertainty . The bias is the deviation of the average k eff of the validation suite from unity. The bias uncertainty accounts for the statistical uncertainty in the bias based on the standard deviation, sample size, and distribution of k eff values of the validation suite. The values of bias and bias uncertainty ensure that the systems predicted to be subcritical by the computational method will indeed be subcritical. The bias and bias uncertainty are often combined to determine an upper subcritical limit (USL) or computational margin that can then be applied to safety analysis calculations. Many methods have been developed by different organizations to calculate the bias and bias uncertainty for various types of criticality analyses. Each of these methods typically requires that the validity of various underpinning statistical assumptions be confirmed to demonstrate that the method is appropriate for the analysis of a given validation suite. An example of the validation decision making flow is shown in Fig.1. As shown in Fig. 1, the analyst performing the validation fits a trend line to the data and performs a test to determine if the trend was a statistically better representation of the data than if it were treated as an uncorrelated sample. If the trend line is a better representation of the data, then the analyst uses any one of a number of trending techniques to determine the bias and bias uncertainty. If a trend is not an appropriate representation of the data, then the analyst proceeds to perform a normality assessment for the data. If the normal assumption can be shown to be acceptable, then the analyst calculates the bias and bias uncertainty with the parametric technique. If the assumption of normality cannot be justified, then the nonparametric technique is used. Once the decision flow has been followed and the appropriate technique has been selected, the bias and bias uncertainty is typically combined with an administrative margin to determine a USL below which calculated values of k eff for safety analysis models can be considered subcritical. The calculations used in each decision are often performed with spreadsheets or with small programs available at various sites performing criticality analyses. Expertise in understanding and interpreting the results must be maintained to perform these calculations. This can often be an error-prone process. Oak Ridge National Laboratory (ORNL) is currently developing the Validation and Data Evaluation Resource (VADER) to simplify and automate the criticality safety validation process and to provide a software quality assurance pedigree to the calculational methods used. This paper discusses the use of the Fulcrum user interface with VADER, the anticipated initial capabilities of VADER to perform validation analyses, and the output from the code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Development and assessment of a reactor system prognosis model with physics-guided machine learning

Autonomous control systems provide recommendations to help operators in decision-making during plant operations ranging from normal operation to accident management. An important step of autonomous control is prognosis. In nuclear engineering domain, prognosis is the process of predicting future conditions of a system or equipment based on present signs and symptoms of a fault. The prognosis model allows predicting future reactor states for possible candidate control strategies so that the outcomes can be evaluated to determine the best control strategy. The prognosis model requires representing direct relationships between the symptoms and the predictions. In nuclear engineering, computational simulations are approximate representations of the operation of the real system. However, prognosis with computational simulations requires high computation power and time due to possible large number of scenarios. Necessary computation resources can be reduced with machine learning (ML) approach for fast predictions by building a surrogate function using the simulation data. A critical issue is, ML models are ignorant of physical knowledge, and these models approximate statistical relationships between the system variables. This ignorance can produce results that are inconsistent with physical laws, even if an optimal result is achieved from a mathematical point of view. Physics-guided machine learning (PGML) is an approach to tackle this issue. Here, this work formulates and illustrates a framework to guide development and assessment of the ML-based prognosis model for autonomous control systems. The development of the prognosis model considers the training of a ML model which consists of optimizing many aspects of the ML approach. The assessment of the prognosis model considers training data limitations and uncertainties of the ML approach. Prognosis models with standalone ML and PGML are developed and assessed on the loss-of-flow scenario of Experimental Breeder Reactor II. The results indicate that PGML based prognosis model has the best performance compared to other prognosis models.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2011

Thermal response of a series- and parallel-connected solar energy storage to multi-day charge sequences

The thermal response of a multi-tank thermal storage was studied under variable charge conditions. Tests were conducted on an experimental apparatus that simulated the thermal charging of the storage system by a solar collector over predetermined (prescribed) daylong periods. The storage was assembled from three standard 270 L hot-water storage tanks each charged through coupled, side-arm, natural convection heat exchangers which were connected in either a series- or parallel-flow configuration. Both energy storage rates and tank temperature profiles were experimentally measured during charge periods representative of two consecutive clear days or combinations of a clear and overcast day. During this time, no draw-offs were conducted. Of particular interest was the effect of rising and falling charge-loop temperatures and collector-loop flow rate on storage tank stratification levels. Results of this study show that the series-connected thermal storage reached high levels of temperature stratification in the storage tanks during periods of rising charge temperatures and also limited destratification during periods of falling charge temperature. This feature is a consequence of the series-connected configuration that allowed sequential stratification to occur in the component tanks and energy to be distributed according to temperature level. This effect was not observed in the parallel charge configuration. A further aspect of the study investigated the effect of increasing charge-loop flow rate on the temperature distribution within the series-connected storage and showed that, at high flow rates, the temperature distributions were found to be similar to those obtained during parallel charging. A disadvantage of both the high-flow series-connected and parallel-connected multi-tank storage is that falling charge-loop temperatures, which normally occur in the afternoon, tend to mix and destratify the storage tanks. (author)

14 SOLAR ENERGY↗

DOE OSTI2014

Staged, High-Pressure Oxy-Combustion Technology: Development and Scale-Up

Washington University in St. Louis and its project partners are developing a unique pressurized oxy-combustion process that aims to improve efficiency and costs by reducing the recycling of flue gas to near zero. Normally, in the absence of recycled flue gas or another inert gas, combustion of fuel and oxygen results in a dramatic increase in temperature of the combustion products and radiant energy, as compared to combustion in air. High heat flux to the boiler tubes may result in a tube surface temperatures that exceed safe operating limits. In the Staged Pressurized Oxy-Combustion (SPOC) process, this problem is addressed by staging the delivery of fuel and by novel combustion design that allows control of heat flux. In addition, the main mode of heat transfer to the steam cycle is by radiation, as opposed to convection. Therefore, the requirement for recycling large amounts of flue gas, for temperature control or to improve convective heat transfer, is eliminated, resulting in a reduction in auxiliary loads. The following report contains a detailed summary of scientific findings and accomplishments for the period of Oct. 1, 2013 to Sept 30, 2014. Results of ASPEN process and CFD modelling activities aimed at improving the SPOC process and boiler design are presented. The effects of combustion pressure and fuel moisture on the plant efficiency are discussed. Combustor pressure is found to have only a minor impact beyond 16 bar. For fuels with moisture content greater than approx 30%, e.g. coal/water slurries, the amount of latent heat of condensation exceeds that which can be utilized in the steam cycle and plant efficiency is reduced significantly. An improved boiler design is presented that achieves a more uniform heat flux profile. In addition, a fundamental study of radiation in high-temperature, high-pressure, particle-laden flows is summarized which provides a more complete understanding of heat transfer in these unusual conditions and to allow for optimization. The results reveal that for the SPOC design, absorption and emission due to particles is the dominant factor for determining the wall heat flux. The mechanism of “radiative trapping” of energy within the high-temperature flame region and the approach to utilizing this mechanism to control wall heat flux are described. This control arises, by design, from the highly non-uniform (non-premixed) combustion characteristics within the SPOC boiler, and the resulting gradients in temperature and particle concentration. Finally, a simple method for estimating the wall heat flux in pressurized combustion systems is presented.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2015

Second-sound studies of coflow and counterflow of superfluid {sup 4}He in channels

We report a comprehensive study of turbulent superfluid {sup 4}He flow through a channel of square cross section. We study for the first time two distinct flow configurations with the same apparatus: coflow (normal and superfluid components move in the same direction), and counterflow (normal and superfluid components move in opposite directions). We realise also a variation of counterflow with the same relative velocity, but where the superfluid component moves while there is no net flow of the normal component through the channel, i.e., pure superflow. We use the second-sound attenuation technique to measure the density of quantised vortex lines in the temperature range 1.2 K ≲ T ≲ T{sub λ} ≈ 2.18 K and for flow velocities from about 1 mm/s up to almost 1 m/s in fully developed turbulence. We find that both the steady-state and temporal decay of the turbulence significantly differ in the three flow configurations, yielding an interesting insight into two-fluid hydrodynamics. In both pure superflow and counterflow, the same scaling of vortex line density with counterflow velocity is observed, L∝V{sub cf}{sup 2}, with a pronounced temperature dependence; in coflow instead, the vortex line density scales with velocity as L ∝ V{sup 3/2} and is temperature independent; we provide theoretical explanations for these observations. Further, we develop a new promising technique to use different second-sound resonant modes to probe the spatial distribution of quantised vortices in the direction perpendicular to the flow. Preliminary measurements indicate that coflow is less homogeneous than counterflow/superflow, with a denser concentration of vortices between the centre of the channel and its walls.

42 ENGINEERING↗

DOE OSTI2015

Regional and Local Geothermal Potential Evaluation: Examples from the Great Basin, USA, Iceland and East Africa

In both tectonically and magmatically mediated geothermal systems, geologic structures play a crucial role in focusing geothermal fluid circulation in the upper crust. In a variety of geothermal provinces worldwide, normal and strike-slip faults, caldera related faults, as well as fractures and fissures related to dikes and other magmatic intrusions generate the fracture permeability utilized for geothermal circulation. Localities where the above structures intersect and interact are characterized by the most robust generation and maintenance of fracture permeability, and therefore are the most favorable conduits for geothermal circulation. We present an evaluation of the structural geometries that tend to generate high density faulting and fracturing and an evaluation of the orientations of those structures with respect to the ambient stress conditions in three regions of high geothermal potential; The Great Basin, USA, Iceland, and east Africa. These analyses provide first-order constraints on areas that have favorable conditions for permeability and where future exploration for geothermal resources could be focused. In the Great Basin, USA, fault step-overs, fault terminations and accommodation zones are among the most common structural settings of known geothermal activity. Analysis of the geometry of these structures with respect to the ambient stress conditions reveals locations where faults associated with these structures are favorably oriented for enhanced fracture permeability. In magmatically active extensional provinces, like Iceland and east Africa, heat associated with shallow magmatic intrusions beneath volcanic centers drives geothermal circulation in near-surface fracture permeability conduits. Accentuated fault and fracture density and fracture permeability is generated by both axially oriented faults and fractures associated with axial rift zones and radially oriented structures near volcanic centers. Although heat from magmatic intrusion is most robust at volcanic centers, a high density of young intrusions may actually result in decreased permeability through closing of fracture networks. The locations of these three types of features; axial faults and fractures, radial structures, and young magmatic intrusions, all play a critical role in determining the most favorable locations for fracture permeability and fluid flow in magmatic geothermal systems. Within and around the margins of volcanic centers, where radial structures associated with volcanic centers interact and intersect with axially oriented structures associated with axial extension are the most favorable locations for fracture permeability and geothermal fluid circulation in magmatic extensional geothermal systems.

15 GEOTHERMAL ENERGY↗

DOE OSTI2020

Certifying the Performance of Fixative Technologies under Open Air Demolition Activities for D and D - 20096

Department of Energy (DOE) facilities undergoing deactivation and decommissioning (D and D) activities struggle with safety concerns of residual radioactive contamination leftover after gross decontamination efforts have concluded. The positive effects of implementing fixative technologies for decommissioning and maintenance is known across the DOE complex. However, there is no standardized metric in place that quantifies the performance of these technologies such as incombustible fixative platforms that immobilize the contamination in a solid polymer material under normal operating conditions or when exposed to stressors. The primary focus of this research is to provide the empirical data necessary to properly characterize the effects of implementing fixative technologies on mitigating the release of contamination. In addition, this effort has been used to identify and establish uniform testing protocols to quantify operational parameters of fixating platforms and properly credit such systems under a variety of operational conditions and stressors that can arise during open air demolition efforts (e.g. impact, water, thermal, etc.). The established protocols will be provided to the ASTM International E10.03 Subcommittee on Radiological Protection for Decontamination and Decommissioning of Nuclear Facilities and Components for the potential of formal standardization. Attributes to be addressed during testing are material compromise from impact stress, thermal stress for incombustible fixative materials, and immobilization factors (how well an immobilizing platform is capable of retaining fixated material during thermal and impact stressors). In developing these test methods, a greater understanding of the material's behavior under anticipated and unanticipated events, such as decommissioning activities and contingency events as outlined in the Basis of Interim Operations (BIO), can be successfully characterized. There are several key components of the experimental methodology that is essential in empirically certifying fixative technologies. A surrogate contaminant with a unique signature has been utilized with the implementation of a uniform contamination process of test coupons that is quantifiable and replicable for direct fixative comparisons. A modular test chamber has also been used to apply the various stressors of interest. Collection of released contamination that includes airborne and resettled contaminant particles and analyzed using mass spectrometry. This research directly supports DOE complex-wide concerns in terms of final disposition of nuclear facilities and the considerations made during their time between ceasing facility operation and final disposition. The results of this effort aims to provide a standardized road map and logical decision flow to better facilitate fixative technologies into D and D activities required to achieve the desired facility end state. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗