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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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DOE OSTI2022

1 kV, 10-kW SiC-Based Quadruple Active Bridge DCX Stage in a DC to Three-Phase AC Module for Medium-Voltage Grid Integration

Interfacing low voltage dc to medium voltage three-phase ac grid is often based on series-stackable modular converter architectures. To minimize energy storage requirements, it is advantageous to employ a quadruple active bridge (QAB) stage operating as a "dc transformer" in each stackable module. The QAB stage offers three isolated dc link voltages, which then allow for flexible stacking of three single-phase dc-to-ac inverter stages. Each of the module phases processes a pulsating power having a component at twice the line frequency. This presents a challenge in maintaining zero voltage switching (ZVS) on the secondary sides of the QAB during low-power portions of the line cycle. This article is focused on the design of the QAB stage. A detailed analysis of ZVS switching waveforms is presented, including effects of nonlinear device capacitances. It is shown how ZVS can be achieved at all times using a relatively small circulating current provided by the magnetizing inductance of the high-frequency transformer. Analytical expressions are given for the optimal values of the magnetizing inductance and the dead times of the QAB primary and secondary bridges. This approach is verified by experimental results on a 1 kV, 10-kW SiC-based prototype, demonstrating a relatively flat efficiency curve with a peak efficiency of 97.1% at 75% load.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2022

Control Design of Series-connected PV-powered Grid-forming Converters via Singular Perturbation

We report modular architectures that consist of several series-connected dc-ac converters have been a focal point of recent innovations in transformerless medium-voltage applications. In this paper, we consider an architecture consisting of dc-ac modules containing a quadruple active bridge dc-dc converter, which generates three floating dc links that feed grid-side dc-ac inverters. Practical implementation of such a converter module in PV systems requires a variety of controllers that collectively achieve maximum power point tracking, dc-link regulation, and ac-side power control. Design of such multi-loop systems is generally quite challenging due to the potential for destabilizing interactions among loops. Here we propose a design approach where singular perturbation theory is used to decompose the timescales at which each control loop operates and provides a systematic framework for parametric selection. Our approach also ensures system stability of multiple modules with identical controls connected in series across a grid. The paper concludes with experimental results of three 1000 W series-connected converter modules across a stiff grid.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2018

Estimating Material Properties Under Extreme Conditions by Using Bayesian Model Calibration with Functional Outputs

Summary Dynamic material properties experiments provide access to the most extreme temperatures and pressures attainable in a laboratory setting; the data from these experiments are often used to improve our understanding of material models at these extreme conditions. We apply Bayesian model calibration to dynamic material property applications where the experimental output is a function: velocity over time. This framework can accommodate more uncertainties and facilitate analysis of new types of experiments relative to techniques traditionally used to analyse dynamic material experiments. However, implementation of Bayesian model calibration requires more sophisticated statistical techniques, because of the functional nature of the output as well as parameter and model discrepancy identifiability. We propose a novel Bayesian model calibration process to simplify and improve the estimation of the material property calibration parameters. Specifically, we propose scaling the likelihood function by an effective sample size rather than modelling the auto-correlation function to accommodate the functional output. Additionally, we propose sensitivity analyses by using the notion of 'modularization' to assess the effect of experiment-specific nuisance input parameters on estimates of the physical parameters. The Bayesian model calibration framework proposed is applied to dynamic compression of tantalum to extreme pressures, and we conclude that the procedure results in simple, fast and valid inferences on the material properties for tantalum.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Synthetic OCP heterodimers are photoactive and recapitulate the fusion of two primitive carotenoproteins in the evolution of cyanobacterial photoprotection

The orange carotenoid protein (OCP) governs photoprotection in the majority of cyanobacteria. It is structurally and functionally modular, comprised of a C-terminal regulatory domain (CTD), an N-terminal effector domain (NTD) and a ketocarotenoid; the chromophore spans the two domains in the ground state and translocates fully into the NTD upon illumination. Using both the canonical OCP1 from Fremyella diplosiphon and the presumably more primitive OCP2 paralog from the same organism, we show that an NTD-CTD heterodimer forms when the domains are expressed as separate polypeptides. The carotenoid is required for the heterodimeric association, assembling an orange complex which is stable in the dark. Both OCP1 and OCP2 heterodimers are photoactive, undergoing light-driven heterodimer dissociation, but differ in their ability to reassociate in darkness, setting the stage for bioengineering photoprotection in cyanobacteria as well as for developing new photoswitches for biotechnology. Additionally, we reveal that homodimeric CTD can bind carotenoid in the absence of NTD, and name this truncated variant the C-terminal domain-like carotenoid protein (CCP). This finding supports the hypothesis that the OCP evolved from an ancient fusion event between genes for two different carotenoid-binding proteins ancestral to the NTD and CTD. We suggest that the CCP and its homologs constitute a new family of carotenoproteins within the NTF2-like superfamily found across all kingdoms of life.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2016

Modeling Hybrid Nuclear Systems With Chilled-Water Storage

Air-conditioning loads during the warmer months of the year are large contributors to an increase in the daily peak electrical demand. Traditionally, utility companies boost output to meet daily cooling load spikes, often using expensive and polluting fossil fuel plants to match the demand. Likewise, heating, ventilation, and air conditioning (HVAC) system components must be sized to meet these peak cooling loads. However, the use of a properly sized stratified chilled-water storage system in conjunction with conventional HVAC system components can shift daily energy peaks from cooling loads to off-peak hours. This process is examined in light of the recent development of small modular nuclear reactors (SMRs). In this paper, primary components of an air-conditioning system with a stratified chilled-water storage tank were modeled in FORTRAN 95. A basic chiller operation criterion was employed. Simulation results confirmed earlier work that the air-conditioning system with thermal energy storage (TES) capabilities not only reduced daily peaks in energy demand due to facility cooling loads but also shifted the energy demand from on-peak to off-peak hours, thereby creating a more flattened total electricity demand profile. Thus, coupling chilled-water storage-supplemented HVAC systems to SMRs is appealing because of the decrease in necessary reactor power cycling, and subsequently reduced associated thermal stresses in reactor system materials, to meet daily fluctuations in cooling demand. Finally and also, such a system can be used as a thermal sink during reactor transients or a buffer due to renewable intermittency in a nuclear hybrid energy system (NHES).

25 ENERGY STORAGE↗
DOE OSTI2021

Advances in engineering microbial biosynthesis of aromatic compounds and related compounds

Abstract Aromatic compounds have broad applications and have been the target of biosynthetic processes for several decades. New biomolecular engineering strategies have been applied to improve production of aromatic compounds in recent years, some of which are expected to set the stage for the next wave of innovations. Here, we will briefly complement existing reviews on microbial production of aromatic compounds by focusing on a few recent trends where considerable work has been performed in the last 5 years. The trends we highlight are pathway modularization and compartmentalization, microbial co-culturing, non-traditional host engineering, aromatic polymer feedstock utilization, engineered ring cleavage, aldehyde stabilization, and biosynthesis of non-standard amino acids. Throughout this review article, we will also touch on unmet opportunities that future research could address.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Power distribution estimation method for SMR using ex-core detectors: experimental demonstration by plural control rod patterns at KUCA

The power distribution estimation method based on the ex-core detectors, or PHOEBE, was demonstrated at Kyoto University Critical Assembly, KUCA. Generally, core monitoring systems use in-core neutron detectors. Since inside the core is a harsh environment, the maintainability and reliability of the detectors are deteriorated. On the other hand, the environment outside the core is milder: core monitoring by the ex-core detectors improves the maintainability and reliability especially for small modular reactors and micro-reactors. However, neutron information from the inner region of the core is lost at the ex-core detectors. To recover the information, the authors proposed to utilize the power correlation between the fuel regions. PHOEBE concept was demonstrated at KUCA with distorted power distributions simulated by control rod patterns. The relative power distribution estimated by PHOEBE agreed with that calculated by Monte Carlo simulation code MVP. PHOEBE approximately reproduced the trend of the distorted power distributions calculated by MVP code. By contrast, the case of without power correlation between fuel regions produced significant different power distributions. Therefore, the advantage of considering the power correlation between fuel regions was also demonstrated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Comparative analysis of energy deposition modes available in Serpent 2 within the framework of the supercritical water reactor - Fuel qualification test reactor physics benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor (SCW-SMR) technology is in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. As a main purpose of the project, proper estimates of energy deposition and its spatial distribution are prerequisites for the accurate analysis of safety related parameters of the SCW-SMR concept under development. A supercritical water reactor fuel computational benchmark model, provided by Canadian Nuclear Laboratories, was applied for detailed comparison of different energy deposition calculation options available in the Serpent 2 Monte Carlo code. The effect of energy deposition options on the normalization of the results as well as on the spatial distribution of the energy deposition are discussed. Consistent energy deposition calculation methods are presented between three Monte Carlo codes, viz., Serpent 2, MCNP6 and OpenMC. Although resource-intensive, the use of the coupled neutron-photon transport mode of Serpent 2 is recommended for accurate spatial and quantitative characterization of energy deposition in the SCW-SMR fuel assemblies, accounting for both neutron and photon heating of all the materials. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Uncertainty Propagation from Experiment Measurements to Modeling Approaches: A Case for SMR Steam Entrainment Testing

To license new and advanced reactor designs, regulators must be convinced that their unique safety cases—relative to existing large scale reactors—have been adequately addressed by the designed reactor protection systems. In water cooled small modular reactors (SMRs), droplet entrainment in steam flow has significant implications on the progression of accident scenarios due to its compact design features, which requires representative test data applicable to SMR designs. Computer code, modeling and simulation (M&S) tools and models require adequate verification, assessment, and qualification. This includes M&S results validation against scaled empirical data within allowable uncertainty bands to gain regulatory approvals during the various stages of reactor system design, demonstration, and commercialization. However, measurement uncertainty within the empirical datasets and test data applicability ranges requires careful consideration of M&S inputs (i.e., boundary conditions, and initial conditions), and verification and validation efforts. This study focuses on uncertainty quantification in designing scaled test facilities for SMR applications with appropriate measurements and a standard data-reduction method to estimate thermal hydraulics characteristics parameters that incorporate physics phenomena of interest. In addition, this study supports the evaluation model development and assessment process using M&S that interfaces with advanced computing tools and digital twin capabilities. This will allow synchronization between experiment and modeling approaches for droplet entrainment testing and analysis, improving diagnostics, prognostics, and decision-making to accelerate regulatory approval.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models

An integrated simulation platform that couples probabilistic analysis-based tools with model-based simulation tools can provide valuable insights for reactive and proactive responses to plant operating conditions. The objective of this work is to demonstrate the benefits of a partial implementation of the Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Framework Specification through the coupling of advanced PRA capabilities and accurate multi-physics plant models. Coupling a probabilistic model with a multi-physics model will aid in design, operations, and safety by providing a more accurate understanding of plant behavior. This represents the first attempt at actually integrating these two types of analyses for a control system used for operations, on a faster than real-time basis. This report documents the development of the basic communication capability to exchange data with the probabilistic model using Reliability Workbench (RWB) and the multi-physics model using Dymola. The communication pathways from injecting a fault (i.e., failing a component) to the probabilistic and multi-physics models were successfully completed. This first version was tested with prototypic models represented in both RWB and Modelica. First, a simple event tree/fault tree (ET/FT) model was created to develop the software code to implement the communication capabilities between the dynamic-link library (dll) and RWB. A program, written in C#, successfully communicates faults to the probabilistic model through the dll. A systems model of the Advanced Liquid-Metal Reactor–Power Reactor Inherently Safe Module (ALMR-PRISM) design developed under another DOE project was upgraded using Dymola to include proper interfaces to allow data exchange with the control application (ConApp). A program, written in C+, successfully communicates faults to the multi-physics model. The results of the example simulation were successfully plotted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Opportunities in SMR Emergency Planning

Using year 2014 cost information gathered from twenty different locations within the current commercial nuclear power station fleet, an assessment was performed concerning compliance costs associated with the offsite emergency Planning Standards contained in 10 CFR 50.47(b). The study was conducted to quantitatively determine the potential cost benefits realized if an emergency planning zone (EPZ) were reduced in size according to the lowered risks expected to accompany small modular reactors (SMR). Licensees are required to provide a technical basis when proposing to reduce the surrounding EPZ size to less than the 10 mile plume exposure and 50 mile ingestion pathway distances currently being used. To assist licensees in assessing the savings that might be associated with such an action, this study established offsite emergency planning costs in connection with four discrete EPZ boundary distances, i.e., site boundary, 2 miles, 5 miles and 10 miles. The boundary selected by the licensee would be based on where EPA Protective Action Guidelines are no longer likely to be exceeded. Additional consideration was directed towards costs associated with reducing the 50 mile ingestion pathway EPZ. The assessment methodology consisted of gathering actual capital costs and annual operating and maintenance costs for offsite emergency planning programs at the surveyed sites, partitioning them according to key predictive factors, and allocating those portions to individual emergency Planning Standards as a function of EPZ size. Two techniques, an offsite population-based approach and an area-based approach, were then employed to calculate the scaling factors which enabled cost projections as a function of EPZ size. Site-specific factors that influenced source data costs, such as the effects of supplemental funding to external state and local agencies for offsite response organization activities, were incorporated into the analysis to the extent those factors could be representatively apportioned.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Status Update of the Majorana Demonstrator Neutrinoless Double Beta Decay Experiment

Neutrinoless double beta decay searches play a major role in determining neutrino properties, in particular the Majorana or Dirac nature of the neutrino and the absolute scale of the neutrino mass. The consequences of these searches go beyond neutrino physics, with implications for Grand Unification and leptogenesis. The Majorana Collaboration is assembling a low-background array of high purity Germanium (HPGe) detectors to search for neutrinoless double-beta decay in 76 Ge. The Majorana Demonstrator, which is currently being constructed and commissioned at the Sanford Underground Research Facility in Lead, South Dakota, will contain 44 kg (30 kg enriched in 76 Ge) of HPGe detectors. Its primary goal is to demonstrate the scalability and background required for a tonne-scale Ge experiment. This is accomplished via a modular design and projected background of less than 3 cnts/tonne-yr in the region of interest. The experiment is currently taking data with the first of its enriched detectors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

IAEA CRP on HTGR Uncertainties in Modeling: Assessment of Phase I Lattice to Core Model Uncertainties

Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented by the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Advanced Nuclear Technology: Integrated Pressurized Water Reactor (iPWR) Containment Aerosol Deposition Behavior

Under assumed severe accident conditions, a nuclear reactor’s containment vessel provides a defense-in-depth function by preventing radionuclide particle release. This is achieved by creating a physical barrier and by decontamination (removal of aerosolized particles produced during the accident). Decontamination occurs through active mechanical systems (where applicable) and passive natural occurring phenomena. Due to their comparatively higher containment surface area to volume ratios when compared to large light water reactors (LWRs), the integrated pressurized water reactor (iPWR) subcategory of small modular reactors (SMRs) has design features that increase the significance of the passive decontamination factors, due to the following natural occurring phenomena: gravitational settling, thermophoresis, diffusiophoresis, and impaction due to convective flows. The purpose of this study is to provide estimates for the decontamination associated with these phenomena. Specifically, these results provide decontamination factors associated with the thermal-hydraulic and geometric parameters that characterize iPWRs, based on the experimental work documented in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Special Purpose Application Reactors: Systems Integration Decision Support

Microreactors, also known as very-Small Modular Reactors (vSMRs) or Special Purpose Reactors (SPRs), are being considered for use in unique applications where other methods of megawatt level energy production are uneconomical or unavailable. For the purposes of this report, a microreactor is defined as meeting the following criteria: • Factory manufacturable • Easily transportable by truck, plane, train, and/or ship • Produce no more than 20 megawatt thermal (MWth) energy, in order to qualify as Hazard Category 2 under 10CFR830 • Maintains neutronic simplicity, allowing semi- or fully-autonomous operation In the United States (U.S.), there are two general types of microreactors under development: high-temperature gas reactors (HTGRs) and heat pipe reactors. These two reactor types may, due to their differences in working fluid temperatures, be more effectively suited to different use cases. Additionally, they will require different interfacing components depending on their intended use. This report identifies a broad range of use cases for micro-reactors. Additionally, multiple reactor technologies are examined with an emphasis on the reactor sub-systems and interconnections. Using this information, a decision framework is developed to guide future efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Calculation of Helium Coolant Behavior in A Single Cooling Channel in MHTGR Reflector Region During Pressurized Conduction Cooldown Scenario Using the COMSOL Multiphysics Code

The objective of the calculation performed and discussed in this report was to examine the behavior of the helium flow through a single cooling channel in the modular high-temperature gas-cooled reactor (MHTGR) core during a pressurized conduction cooling scenario (PCC). The region of interest for the single cooling channel flow calculation is in the upper reflector blocks. Flow from the upper reflector blocks, during the PCC scenario, moves upward and into the upper plenum.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗