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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Johnson Noise Thermometry for Advanced Small Modular Reactors

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

Britton, Jr, Charles L.↗
DOE OSTI2012

Digial Technology Qualification Task 2 - Suitability of Digital Alternatives to Analog Sensors and Actuators

The next generation reactors in the U.S. are an opportunity for vendors to build new reactor technology with advanced Instrumentation and Control Systems (control rooms, DCS, etc.). The advances made in the development of many current generation operating reactors in other parts of the world are being used in the design and construction of new plants. These new plants are expected to have fully integrated digital control rooms, computerized procedures, integrated surveillance testing with on-line monitoring and a major effort toward improving the O&M and fault survivability of the overall systems. In addition the designs are also incorporating major improvements in the man-machine interface based on lessons learned in nuclear and other industries. The above relates primarily to the scope of supply in instrumentation and control systems addressed by Chapter 7 of the Standard Review Plan (SRP) NUREG-0800 (Reference 9.5), and the associated Balance of Plant (BOP) I&C systems. This does not relate directly to the actuator and motor, breaker, initiation circuitry, valve position, etc. which is the subject of this report and normally outside of the traditional Distributed Control System (DCS), for both safety and non-safety systems. The recommendations presented in this report will be used as input to I&C research programming for the implementation of lessons learned during the early phases of new build both for large light water reactors (LWR) and also small modular reactors (SMR). This report is intended to support current research plans and provide user (vendor, owner-operator) input to the optimization of these research plans.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2012

EngenuitySC Commercialization and Entrepreneurial Training Project

A team led by EngenuitySC has performed education and outreach on development of advanced energy markets that will enable wider use of clean energy technologies. This report details the efforts that have made significant advances to improve the market place through education, outreach, and increased communications between industry members. The project resulted in two self-funded industry clusters known as the Fuel Cell Collaborative and NuHub. This project has focused on building and strengthening the leading clean energy clusters in South Carolina: nuclear energy and fuel cell technologies. For the nuclear industry, a new cluster was developed that is now known as NuHub. This cluster has already engaged over 25 nuclear industry leaders or suppliers, four public sector partners, six community economic development foundations, and nearly ten academic partners in a 175 mile radius between Augusta, Georgia and Charlotte, North Carolina. Our outreach has touched over 2,000 stakeholders through the website alone, not including the public audiences and members of the business community reached through news stories and releases that were distributed to over 620 print and online publications. NuHub has established a formal leadership structure, developed subcommittees to focus on industry issues, instituted educational programs for the workforce, and created an industry funding structure that will sustain the industry cluster and mission. NuHub has participated in a wide-variety of community building and outreach activities since its formation under this grant. In the two years since its creation in 2010, we have initiated efforts focused in four main areas that correlate with the four NuHub subcommittees including: innovation, workforce development, industry engagement, and marketing and communications. NuHub successfully raised over $\$$160,000 in both public and private funding, which has supported work to grow the cluster and engage partners including NuScale, Fluor, and Holtec International for research about deployment of advanced small modular reactor (SMR) technologies. The workforce training efforts from NuHub have focused on assisting existing industry to fill positions needed to construct and operate new nuclear plants being built at the VC Summer plant in Jenkinsville, SC and at Plant Votgle in Augusta, Georgia both of whom are constructing the first nuclear reactors (Westinghouse AP 1000 units), to be built in over 30 years. This includes a partnership with Midlands Technical College to train reactor operators and the development of training facilities to support workforce development activities. It is anticipated that approximately 70 students a year will be trained through these programs in the next five years, and it will be expanded to meet new industry needs.

08 HYDROGEN↗
DOE OSTI2013

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

This report describes the status of ongoing research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Johnson Noise Thermometry System Requirements

This document is intended to capture the requirements for the architecture of the developmental electronics for the ORNL-lead drift-free Johnson Noise Thermometry (JNT) project conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced Small Modular Reactor (SMR) Research and Development (R&D) program. The requirements include not only the performance of the system but also the allowable measurement environment of the probe and the allowable physical environment of the associated electronics. A more extensive project background including the project rationale is available in the initial project report [1].

Britton, Jr., Charles L.↗
DOE OSTI2013

PDRD (SR13046) Tritium Production Final Report

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

Baseline Concept Description of a Small Modular High Temperature Reactor

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities. The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

This report describes research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Baseline Concept Description of a Small Modular High Temperature Reactor

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Depth-dependent Vertical-to-Horizontal (V/H) Ratios of Free-Field Ground Motion Response Spectra for Deeply Embedded Nuclear Structures

This report documents the results of a study to determine the depth-dependent V/H ratios of ground motion response spectra in the free field. The V/H ratios reported herein were developed from a worldwide database of surface and downhole acceleration recordings obtained from 45 vertical array stations. This database was specifically compiled for this project, and includes information from a diversity of active tectonic regions (California, Alaska, Taiwan, Japan), site conditions (rock to soft soil), ground motion intensity levels (PGAs between 0.01 g and 0.50 g), magnitudes (between ML 2.78 and JMA 8.1), epicentral distances (between 3.2 km and 812 km), and source depths (between 1.2 km and 112 km), as well as sensors at surface and at a wide range of depths relevant to the project. To study the significance of the depth effect, V/H ratios from all the records were sorted into a number of depth bins relevant to the project, and statistics (average, standard deviation, coefficient of variation, 16th, 50th, and 84th percentiles) of the V/H ratios within each bin were computed. Similar analyses were repeated, controlling for different site conditions, ground motion intensity levels, array locations, and source depths, to study their relative effect on the V/H ratios. Our findings confirm the importance of the depth effect on the V/H ratios. The research findings in this report can be used to provide guidance on the significance of the depth effect, and the extent to which this effect should be considered in the seismic design of deeply embedded SMR structures and NPP structures in general.

42 ENGINEERING↗
DOE OSTI2015

Critical Heat Flux Phenomena at High Pressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m 2 s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Baseline Concept Description of a Small Modular High Temperature Reactor

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Preliminary Development of a Work Breakdown Structure (WBS) for Small Modular Reactors (SMRs)

In summary, this preliminary WBS serves as an initial basis for the capital cost component of the economic analysis of SMRs. This preliminary WBS comes from the known WBS for existing, large nuclear power plants and develops the methodology for accounting for the anticipated differences between the current large plants and the projected SMR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Nuclear Security for Floating Nuclear Power Plants

Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology are proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Development of an Automated Decision-Making Tool for Supervisory Control System

This technical report was generated as a product of the Supervisory Control for Multi-Modular Small Modular Reactor (SMR) Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (AdvSMR) Research and Development Program of the US Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular AdvSMR plants. This research activity advances the state of the art by incorporating real-time, probabilistic-based decision-making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides background information on the state of the art of automated decision-making, including the description of existing methodologies. It then presents a description of a generalized decision-making framework, upon which the supervisory control decision-making algorithm is based. The probabilistic portion of automated decision-making is demonstrated through a simple hydraulic loop example.

42 ENGINEERING↗
DOE OSTI2017

CO 2 Capture by Cold Membrane Operation with Actual Power Plant Flue Gas

The main objective of the project was to develop a post-combustion CO 2 capture process based on the hybrid cold temperature membrane operation. The CO 2 in the flue gas from coal fired power plant is pre-concentrated to >60% CO 2 in the first stage membrane operation followed by further liquefaction of permeate stream to achieve >99% CO 2 purity. The aim of the project was based on DOE program goal of 90% CO 2 capture with >95% CO 2 purity from Pulverized Coal (PC) fired power plants with $40/tonne of carbon capture cost by 2025. The project moves the technology from TRL 4 to TRL 5. The project involved optimization of Air Liquide commercial 12” PI-1 bundle to improve the bundle productivity by >30% compared to the previous baseline (DE-FE0004278) using computational fluid dynamics (CFD) modeling and bundle testing with synthetic flue gas at 0.1 MWe bench scale skid located at Delaware Research and Technology Center (DRTC). In parallel, the next generation polyimide based novel PI-2 membrane was developed with 10 times CO 2 permeance compared to the commercial PI-1 membrane. The novel PI-2 membrane was scaled from mini-permeator to 1” permeator and 1” bundle for testing. Bundle development was conducted with a Development Spin Unit (DSU) installed at MEDAL. Air Liquide’s cold membrane technology was demonstrated with real coal fired flue gas at the National Carbon Capture Center (NCCC) with a 0.3 MWe field-test unit (FTU). The FTU was designed to incorporate testing of two PI-1 commercial membrane bundles (12” or 6” diameter) in parallel or series. A slip stream was sent to the next generation PI-2 membrane for testing with real flue gas. The system exceeded performance targets with stable PI-1 membrane operation for over 500 hours of single bundle, steady state testing. The 12” PI-1 bundle exceeded the productivity target by achieving ~600 Nm3/hr, where the target was set at ~455 Nm3/hr at 90% capture rate. The cost of 90% CO 2 capture from a 550 MWe net coal power plant was estimated between 40 and $45/tonne. A 6” PI-1 bundle exhibited superior bundle performance compared to the 12” PI-1 bundle. However, the carbon capture cost was not lower with the 6” PI-1 bundle due to the higher bundle installed cost. A 1” PI-1 bundle was tested to compare bundles with different length / diameter ratios. This bundle exhibited the lowest performance due to the different fiber winding pattern and increased bundle non-ideality. Several long-term and parametric tests were conducted with 3,200 hours of total run-time at NCCC. Finally, the new PI-2 membrane fiber was tested at a small scale (1” modules) in real flue gas and exhibited up to 10 times the CO 2 permeance and slightly lower CO 2 /N 2 selectivity as the commercial PI-1 fiber. This corresponded to a projected 4 - 5 times increase in the productivity per bundle and a potential cost reduction of $3/tonne for CO2 capture, as compared with PI-1. An analytical campaign was conducted to trace different impurities such as NOx, mercury, Arsenic, Selenium in gas and liquid samples through the carbon capture system. An Environmental, Health and Safety (EH&S) analysis was completed to estimate emissions from a 550 MWe net power plant with carbon capture using cold membrane. A preliminary design and cost analysis was completed for 550 tpd (~25 MWe) plant to assess the capital investment and carbon capture cost for PI-1 and PI-2 membrane solutions from coal fired flue gas. A comparison was made with an amine based solution with significant cost advantage for the membrane at this scale. Additional preliminary design and cost analysis was completed between coal, natural gas and SMR flue gas for carbon capture at 550 tpd (~25 MWe) plant.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2017

One Step Hydrogen Generation Through Sorption Enhanced Reforming

One-step hydrogen generation, using Sorption Enhanced Reforming (SER) technology, is an innovative means of providing critical energy and environmental improvements to US manufacturing processes. The Gas Technology Institute (GTI) is developing a Compact Hydrogen Generator (CHG) process, based on SER technology, which successfully integrates previously independent process steps, achieves superior energy efficiency by lowering reaction temperatures, and provides pathways to doubling energy productivity with less environmental pollution. GTI’s prior CHG process development efforts have culminated in an operational pilot plant. During the initial pilot testing, GTI identified two operating risks- 1) catalyst coating with calcium aluminate compounds, 2) limited solids handling of the sorbent. Under this contract GTI evaluated alternative materials (one catalyst and two sorbents) to mitigate both risks. The alternate catalyst met performance targets and did not experience coating with calcium aluminate compounds of any kind. The alternate sorbent materials demonstrated viable operation, with one material enabling a three-fold increase in sorbent flow. The testing also demonstrated operation at 90% of its rated capacity. Lastly, a carbon dioxide co-production study was performed to assess the advantage of the solid phase separation of carbon dioxide- inherent in the CHG process. Approximately 70% lower capital cost is achievable compared to SMR-based hydrogen production with CO2 capture, as well as improved operating costs.

08 HYDROGEN↗