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All sources NASA NTRS DOE OSTI

DOE OSTI2021

Polysulfide-Permanganate Flow Battery Using Abundant Active Materials

A new flow battery is presented using the abundant and inexpensive active material pairs permanganate/manganate and disulfide/tetrasulfide. A wetted material set is identified for compatibility with the strongly oxidizing manganese couple at ambient and elevated temperatures. Both solutions allow high active material solubility, with cells tested at theoretical energy densities up to 43 Wh l −1 for the ∼1.2 V cell. Full cells built with nickel foam electrodes and sodium-exchanged Nafion 115 membranes deliver a baseline area-specific resistance of 2.7 Ω-cm 2 . Incorporation of high-surface-area cobalt-coated carbon paper and high-surface-area stainless steel mesh electrodes, and an expanded Nafion 115 membrane delivers cells with 44% lower resistance at 1.6 Ω-cm 2 . All cells show performance decay over the course of cycling. The Co-decorated carbon paper electrodes provide significant kinetic improvements, shifting electrode performance from non-linear with Ni-foam to linear with a volume-normalized exchange current density value of 3.2 A cm −3 . The expanded membrane provides increased conductivity over the 13 mS cm −1 conductivity observed in as-received, sodium-exchanged Nafion 115. Although boiled membranes provide improved conductivity, it is at the cost of decreased Coulombic efficiency and poorer manufacturability. Full cell models suggest that similar cell resistances (1.7 Ω-cm 2 ) should be feasible with as-received Nafion 115 and advanced electrodes.

25 ENERGY STORAGE↗

DOE OSTI2015

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Geysers Valley CO 2 Cycling Geological Engine (Kamchatka, Russia)

1941–2017 period of the Valley of Geysers monitoring (Kamchatka, Kronotsky Reserve) reveals a very dynamic geyser behavior under natural state conditions: significant changes of IBE (interval between eruptions) and power of eruptions, chloride and other chemical components, and preeruption bottom temperature. Nevertheless, the total deep thermal water discharge remains relatively stable; thus all of the changes are caused by redistribution of the thermal discharge due to giant landslide of June 3, 2007, mudflow of Jan. 3, 2014, and other events of geothermal caprock erosion and water injection into the geothermal reservoir. In some cases, water chemistry and isotope data point to local meteoric water influx into the geothermal reservoir and geysers conduits. TOUGHREACT V.3 modeling of Velikan geyser chemical history confirms 20% dilution of deep recharge water and CO 2 components after 2014. Temperature logging in geysers Velikan (1994, 2007, 2015, 2016, and 2017) and Bolshoy (2015, 2016, and 2017) conduits shows preeruption temperatures below boiling at corresponding hydrostatic pressure, which means partial pressure of CO 2 creates gas-lift upflow conditions in geyser conduits. Velikan geyser IBE history explained in terms of gradual CO 2 recharge decline (1941–2013), followed by CO 2 recharge significant dilution after the mudflow of Jan. 3, 2014, also reshaped geyser conduit and diminished its power.

58 GEOSCIENCES↗

DOE OSTI2018

FEHM V3.0 - V3.X

The numerical background of the FEHM computer code can be traced to the early 1970s when it was used to simulate geothermal and hot dry rock reservoirs. The primary use over a number of years was to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the potential Yucca Mountain repository. Today FEHM is used to simulate groundwater and contaminant flow and transport in deep and shallow, fractured and un-fractured porous media throughout the US DOE complex. FEHM has proved to be a valuable asset on a variety of projects of national interest including Environmental Remediation of the Nevada Test Site, the LANL Groundwater Protection Program, geologic CO2 sequestration, Enhanced Geothermal Energy (EGS) programs, Oil and Gas production, Nuclear Waste Isolation, and Arctic Permafrost. Subsurface physics has ranged from single fluid/single phase fluid flow when simulating basin scale groundwater aquifers to complex multifluid/ multi-phase fluid flow that includes phase change with boiling and condensing in applications such as unsaturated zone surrounding nuclear waste storage facility or leakage of CO2/brine through faults or wellbores. The numerical method used in FEHM is the control volume method (CV) for fluid flow and heat transfer equations which allows FEHM to exactly enforce energy/mass conservation; while an option is available to use the finite element (FE) method for displacement equations to obtain more accurate stress calculations. In addition to these standard methods, an option to use FE for flow is available, as well as a simple Finite Difference scheme.

Pawar, Rajesh↗

DOE OSTI2017

LH2 Transfer Simulation

LH2TS enables to calculate the wasted H2 due to transfer and boil-off while using LH2 as a fuel for transportation. The supply chain from the liquefaction plant to the station can be modeled using real gas equations of state from REFPROP and heat transfer models.

Petitpas, Guillaume↗

DOE OSTI2018

Cryogenic H2 vehicle simulation

CryoH2Vehicle enables to simulate the variations of the thermodynamic states of H2 inside nn onboard vessel, that undergoes n given duty cycle (park/drive/fill). Typical states of charge and boil-off losses can thus be estimated.

Petitpas, Guillaume↗

DOE OSTI2026

sandialabs/P2MAT

P2MAT - A python based user interface to predict melting point and boiling point of chemical compounds.

Kamruzzaman, Methun [Sandia National Lab. (SNL-CA)↗

DOE OSTI2024

Chemical-specific Parameters Dataset

The chemical-specific parameters dataset is searchable for physicochemical information for multiple chemicals simultaneously. After selecting chemicals of interest and the desired parameters, the RAIS will generate a table containing the values, chosen according to an established hierarchy. Results can be downloaded in Excel format. Over 40 parameters are available, including melting point, boiling point, density, density, vapor pressure, water solubility, and Henry’s Law constants. Thirteen primary sources are used to populate the dataset of chemical-specific parameters. These values should be used in cancer risk and noncancer hazard assessments for the calculation of preliminary remediation goals (PRGs), hazard characterization, and transport modeling. Users can select up to 1000 chemicals per query. The dataset supports environmental risk assessments, regulatory decision-making, and environmental planning with tools for benchmarking against risk-based standards. This structured approach ensures a robust evaluation of environmental risks tailored to regulatory needs.

Dolislager, Fred [Oak Ridge National Laboratory (O↗

DOE OSTI2016

Evolution and current understanding of physicochemical characterization of particulate matter from reactivity controlled compression ignition combustion on a multicylinder light-duty engine

Low-temperature compression ignition combustion can result in nearly smokeless combustion, as indicated by a smoke meter or other forms of soot measurement that rely on absorbance due to elemental carbon content. Highly premixed low-temperature combustion modes do not form particulate matter in the traditional pathways seen with conventional diesel combustion. Previous research into reactivity controlled compression ignition particulate matter has shown, despite a near zero smoke number, significant mass can be collected on filter media used for particulate matter certification measurement. In addition, particulate matter size distributions reveal that a fraction of the particles survive heated double-dilution conditions. This paper summarizes research completed at Oak Ridge National Laboratory to date on characterizing the nature, chemistry and aftertreatment considerations of reactivity controlled compression ignition particulate matter and presents new research highlighting the importance of injection strategy and fuel composition on reactivity controlled compression ignition particulate matter formation. Particle size measurements and the transmission electron microscopy results do show the presence of soot particles; however, the elemental carbon fraction was, in many cases, within the uncertainty of the thermal–optical measurement. Particulate matter emitted during reactivity controlled compression ignition operation was also collected with a novel sampling technique and analyzed by thermal desorption or pyrolysis gas chromatography mass spectroscopy. Particulate matter speciation results indicated that the high boiling range of diesel hydrocarbons was likely responsible for the particulate matter mass captured on the filter media. Finally, to investigate potential fuel chemistry effects, either ethanol or biodiesel were incorporated to assess whether oxygenated fuels may enhance particle emission reduction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Investigations into EGR dilution tolerance in a pre-chamber ignited GDI engine

EGR is a useful means to improve fuel economy in spark-ignited gasoline engines, especially under part-load conditions by means of reducing throttling losses. There is also a significant reduction in NOx emissions due to the reduced peak cylinder temperatures. However, the ability of conventional spark ignition systems to reliably ignite dilute mixtures limits the dilution tolerance due to the onset of combustion instability. In this study, an active pre-chamber system was evaluated at part load condition with various EGR rates using a regular grade full boiling range gasoline. A premixed air-fuel mixture was supplied to the pre-chamber, which assists in scavenging the residual mass fractions to enable reliable ignition of EGR dilute mixtures in the main combustion chamber. A range of pre-chamber injection, air-fuel mixture formulation, and ignition timing strategies were evaluated during the EGR sweeps. In addition, 0-D and 1-D simulations were conducted in GT-Power to assess the thermodynamic state and composition in the pre-chamber, as well as estimate the jet momentum at the various dilution rates. The active pre-chamber extended the EGR dilution limit from 20% for conventional SI to above 30%, with a higher pre-chamber flow rate (more scavenging) resulting in increased combustion stability. A tradeoff between the jet momentum and spark timing was observed when analyzing the GT-Power simulations and the pre-chamber jet momentum was found to decrease with increasing EGR rate. At unstable high EGR operation, no relation could be found between the pre-chamber indicated data and main chamber combustion stability.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2026

Hazard and risk analysis framework for nuclear power plant–based integrated energy systems

Employing integrated energy systems (IESs) with nuclear power plants (NPPs) can improve NPP utilization by leveraging dedicated thermal and electric power delivery, but it may also increase operational safety risks. This paper presents a framework to identify and quantify hazards and risks for such IESs. The framework combines accidentology to review past industrial accidents with failure modes and effects analysis (FMEA) to identify potential future incidents. Hydrogen explosion and toxic chemical release hazards are of particular concern. Explosion consequences are quantified using the Bauwens-Dorofeev (Bauwens) and trinitrotoluene equivalent mass (TNT-EM) methods, while chemical release consequences are computed using the Gaussian atmospheric dispersion method. Operational disturbances from direct electrical and thermal integration that may affect NPP safety are modeled using probabilistic risk analysis (PRA). Hazards and risks are then evaluated for regulatory compliance. The framework is applied to IESs comprising pressurized or boiling water reactors supplying three levels of thermal and electrical power to industrial customers. Case studies include high-temperature steam electrolysis hydrogen plants of varying capacities and a synthetic fuel production plant. Sensitivity analysis examines piping component failures in the PRA model as a precursor to cost estimation for thermal extraction line design. Additionally, Fussel-Vessely (FV) and risk increase importance (RII) measures identify risk-informed design improvements for the thermal extraction system. FMEA highlights hazards such as loss of offsite power, prompt loss of electrical load, loss of thermal output, and immediate steam diversion, in addition to hydrogen explosions and toxic chemical releases. Both Bauwens and TNT-EM methods suggest maintaining several hundred meters of separation between the NPP and hydrogen facility to mitigate explosion risks. PRA results show a maximum initiating event frequency increase of 1.15% and an overall risk increase of 0.28%. Importance measure analysis identifies upstream pipe leak isolation components as critical. Evaluating the results against safety regulations, it is concluded that hazards and risks can be managed to comply with regulations through risk-informed thermal and electrical connection designs, component selection, maintenance programs, and safe separation distances between NPPs and integrated industrial facilities.

08 - HYDROGEN↗

DOE OSTI2015

Evaluation of Cooling Conditions for a High Heat Flux Testing Facility Based on Plasma-Arc Lamps

The new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL) uses an infrared plasma-arc lamp (PAL) to deliver incident heat fluxes as high as 27 MW/m 2 . The facility is being used to test irradiated plasma-facing component materials as part of the joint US-Japan PHENIX program. The irradiated samples are to be mounted on molybdenum sample holders attached to a water-cooled copper rod. Depending on the size and geometry of samples, several sample holders and copper rod configurations have been fabricated and tested. As a part of the effort to design sample holders compatible with the high heat flux (HHF) testing to be conducted at the IMTS facility, numerical simulations have been performed for two different water-cooled sample holder designs using the ANSYS FLUENT 14.0 commercial computational fluid dynamics (CFD) software package. The primary objective of this work is to evaluate the cooling capability of different sample holder designs, i.e. to estimate their maximum allowable incident heat flux values. 2D axisymmetric numerical simulations are performed using the realizable k-ε turbulence model and the RPI nucleate boiling model within ANSYS FLUENT 14.0. The results of the numerical model were compared against the experimental data for two sample holder designs tested in the IMTS facility. The model has been used to parametrically evaluate the effect of various operational parameters on the predicted temperature distributions. The results were used to identify the limiting parameter for safe operation of the two sample holders and the associated peak heat flux limits. The results of this investigation will help guide the development of new sample holder designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2014

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (k eff ) Predictions

One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (k eff ) evaluations based on best-available data and methods and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate k eff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the application model. Finally, this paper provides a detailed description of the approach and its technical bases, describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models, and provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Water Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Lastly, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, the capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

CFD Analysis and Design of Detailed Target Configurations for an Accelerator-Driven Subcritical System

High-fidelity analysis has been utilized in the design of beam target options for an accelerator driven subcritical system. Designs featuring stacks of plates with square cross section have been investigated for both tungsten and uranium target materials. The presented work includes the first thermal-hydraulic simulations of the full, detailed target geometry. The innovative target cooling manifold design features many regions with complex flow features, including 90 bends and merging jets, which necessitate three-dimensional fluid simulations. These were performed using the commercial computational fluid dynamics code STAR-CCM+. Conjugate heat transfer was modeled between the plates, cladding, manifold structure, and fluid. Steady-state simulations were performed but lacked good residual convergence. Unsteady simulations were then performed, which converged well and demonstrated that flow instability existed in the lower portion of the manifold. It was established that the flow instability had little effect on the peak plate temperatures, which were well below the melting point. The estimated plate surface temperatures and target region pressure were shown to provide sufficient margin to subcooled boiling for standard operating conditions. This demonstrated the safety of both potential target configurations during normal operation.

43 PARTICLE ACCELERATORS↗

DOE OSTI2016

United States Department of Energy severe accident research following the Fukushima Daiichi accidents

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCOR results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U. S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. Finally, this paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2022

Dancoff-based Wigner-Seitz approximation for the subgroup resonance self-shielding in the VERA neutronic simulator MPACT

The MPACT neutronics module of the Virtual Environment for Reactor Analysis (VERA) has used the subgroup method for resonance self-shielding calculation, for which two-dimensional (2D) fixed-source transport calculations are performed using the method of characteristics for resonance energy groups. When considering thermal feedbacks, the subgroup calculation must be performed at each outer iteration. Therefore, the computing time for cross section processing is a significant burden for computational efficiency. The Dancoff-based Wigner-Seitz approximation (DWA) capability has been implemented into MPACT in conjunction with the subgroup method, which has been used in SCALE/XSProc since SCALE version 6.0 and has recently been called an equivalent Dancoff-factor cell (EDC) method. The issue of relatively large reactivity bias in DWA for the gadolinia rods was resolved by introducing multiple Dancoff factors. Benchmark results for the VERA pressurized and boiling water reactor benchmark suites show that the DWA capability would significantly enhance computational efficiency with comparable accuracy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗