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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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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Creation of Multiple Effect Evaporator and Combined Cycle Modelica Modules, and Optimization of Potable Water Generation from Saltwater Sources (Final Technical Report)

Modelica models for a Heat Recovery Steam Generator (HRSG) and Multiple Effect Evaporator have been developed. These models have been validated against results from literature and from an actual cogeneration system on the campus of NC State University. The models are capable of connecting to other components to create multi-input, multi-output energy systems with storage, also known as integrated energy systems. An existing 35 MW gas turbine model is connected to the HRSG model created for this project to produce 9 bar steam. The high pressure steam feeds a backpressure steam turbine from the library to generate additional electric power. 2 bar steam leaves the turbine to supply an 8 effect MEE to generate water for a community. As part of the effort, three research papers were submitted (two published in ANS Proceedings and one submitted to Applied Energy).

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2016

Analyzing the Impact of Solar Power on Multi-Hourly Thermal Generator Ramping

Solar power generation, unlike conventional forms of electricity generation, has higher variability and uncertainty in its output because solar plant output is strongly impacted by weather. As the penetration rate of solar capacity increases, grid operators are increasingly concerned about accommodating the increased variability and uncertainty that solar power provides. This paper illustrates the impacts of increasing solar power penetration on the ramping of conventional electricity generators by simulating the operation of the Independent System Operator -- New England power system. A production cost model was used to simulate the power system under five different scenarios, one without solar power and four with increasing solar power penetrations up to 18%, in terms of annual energy. The impact of solar power is analyzed on six different temporal intervals, including hourly and multi-hourly (2- to 6-hour) ramping. The results show how the integration of solar power increases the 1- to 6-hour ramping events of the net load (electric load minus solar power). The study also analyzes the impact of solar power on the distribution of multi-hourly ramping events of fossil-fueled generators and shows increasing 1- to 6-hour ramping events for all different generators. Generators with higher ramp rates such as gas and oil turbine and internal combustion engine generators increased their ramping events by 200% to 280%. For other generator types--including gas combined-cycle generators, coal steam turbine generators, and gas and oil steam turbine generators--more and higher ramping events occurred as well for higher solar power penetration levels.

14 SOLAR ENERGY↗
DOE OSTI2016

Experimental validation on the operational and cooling performance of the passive safety system with Atlas

After the Fukushima accident, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions. In Korea, a passive auxiliary feedwater system (PAFS) was developed and adopted to a GEN-III+ nuclear power plant, APR+ to completely replace the conventional auxiliary feedwater system. The integral effect tests were performed to validate the cooling and operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. The representative design basis accidents (DBAs) such as a feedwater line break, a steam line break, and a steam generator tube rupture transients were experimentally simulated in this study. In addition, an integral effect test was performed for simulating a station blackout transient as a limiting beyond DBA. From the integral effect test results, it could be concluded that the APR+ has the capability of coping with the hypothetical DBA and bDBA by adopting PAFS and proper set-points of its operation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1978

Waste heat utilization in industrial processes

A survey is given of new developments in heat exchangers and heat pumps. With respect to practical applications, internal criteria for plant operation are discussed. Possibilities of government support are pointed out. Waste heat steam generators and waste heat aggregates for hot water generation or in some cases for steam superheating are used. The possibilities of utilization can be classified according to the economic improvements and according to their process applications, for example, gascooling. Examples are presented for a large variety of applications.

Weichsel, M.↗
DOE OSTI2019

OSU High Temperature Test Facility Design Technical Report, Revision 2

The HTTF is an integral test facility, which has been designed to model the behavior of interest for a MHTGR during the Depressurized Conduction Cooldown (DCC) event. It also has the potential to conduct limited explorations into the progression of the Pressurized Conduction Cooldown (PCC) event and phenomena during normal operations. The facility is scaled 1:4 by height and 1:64 by volume. A scaling report has been developed and provides further detail into the scaling analysis for the HTTF. The maximum core power for the facility is approximately 2.2MW. The core region will be of a modular design to allow for both the prismatic block core and the pebble bed core designs to be tested. However, only the prismatic block design is discussed in this report. The HTTF consists of the following systems: A primary system consisting of the reactor pressure vessel (RPV), the circulator, cross over ducts and primary piping. The core internals sit inside the RPV and provide support for the core and core heaters. The purpose of the primary system is to maintain the pressure boundary, remove heat from the core through convection, conduction and radiation, and provide heat to the core through a series of electrical heaters. A secondary system consisting of a steam generator, feed water system and main steam system. The purpose of the secondary system is to remove heat from the primary system during heat up, cool down and normal operations. A reactor cavity simulation system consisting of a cavity tank, gas conditioning system and a series of break valves. Note that the prototypical MHTGR sits inside of a cavity. In the event of a break the helium will blowdown into this cavity. At the HTTF, the cavity is modeled as a separate tank, the reactor cavity simulation tank (RCST) and is connected to the RPV through break valves—hot-leg, cold-leg, upper break, and lower break. A reactor cavity cooling system (RCCS) consisting of a series of water cooled panels and the water supply system for these panels. The HTTF RCCS is not a scaled prototypical RCCS system for the MHTGR. It is provided in the HTTF in order to set the boundary condition for radiation heat transfer from the RPV. Various auxiliary systems including the city water supply system, the electrical power supply and distribution system, the chilled water system and the instrument air system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

The Modeling of Synfuel Production Process: ASPEN Model of FT production with electricity demand provided at LWR scale

Synfuels, or electro-fuels (e-fuels) have the unique potential to significantly reduce greenhouse gas (GHG) emissions across the transportation sector. This is especially true for applications with substantial payloads and daily miles traveled, such as long-haul heavy-duty vehicles, rail locomotives, marine vessels and aviation aircrafts that are challenging to directly electrify via battery or fuel cell powertrain technologies. Synfuels, or electro-diesel/electro-jet fuels, have similar properties with the incumbent petroleum fuels, compatible with current infrastructure but have much lower GHG emissions relative to the petroleum counterpart, because they utilize waste carbon dioxide (CO2) streams and green hydrogen (H2) sourced from electrolysis. To achieve substantial reductions in GHG emissions, electricity sources must be zero carbon or near-zero carbon, which is the case with solar, wind, hydro and nuclear power. Compared to the intermittency of solar, wind and hydro, nuclear energy provides a steady energy source. In addition, it’s advantageous for nuclear power to produce synfuels because it provides not only near-zero carbon electricity to displace grid electricity, but also near-zero carbon steam to displace carbon-intensive natural gas combustion for steam generation. The availability of electricity and steam also enables more efficient green hydrogen production by using high-temperature electrolysis. In this work, Argonne National Laboratory (ANL) models a synfuel production process via the Fischer- Tropsch (FT) reaction by using nuclear power to provide electricity and steam. In 2021, using ASPEN Plus software, ANL established a detailed process model of a stand-alone FT production facility, assuming feedstocks of pure CO2 and H2. This stand-alone model can be expanded to integrate H2 production from nuclear power via low-temperature and high-temperature electrolysis at light-water reactor (LWR) scale. This report summarizes the stand-alone ASPEN Plus model results with a detailed mass and energy analysis. Our modeled facility produces 351 MT/day (130,000 gal/day) of FT fuel (a mixture of naphtha, jet fuel, and diesel) by converting 223 MT/day of H2 and 2,387 MT/day of CO2. The FT fuel production energy efficiency is 58% and the carbon conversion efficiency (from CO2 to FT fuel) is 46%. The production of green hydrogen requires 390–470 MWe of electricity, which is compared with the capacity of an LWR plant. For the stand-alone FT process, the detailed energy demand (electricity and heat) is summarized in the table below. Based on the energy supply source and the required temperature, potential insertion points of nuclear energy are identified. Based on the potential nuclear energy utilization, this report discusses potential modification options for expanding the system boundary to integrate nuclear power use, for example on-site hydrogen production via water electrolysis. Modeling of the integrated system is conducted by closely working with ANL and Idaho National Laboratory (INL) collaborators to harmonize design parameters of nuclear plants and the FT production process.

Zang, Guiyan↗
DOE OSTI2017

Improved Safety Injection Flow Map Associated with Target RCS Depressurization to Maintain Core Coolability of OPR1000

In Korea, Severe Accident Management Guidance (SAMG) for Optimized Power Reactor (OPR1000) has been developed through the research results related to the WOG SAMG, in which the operation phase shifts from the Emergency Operation Procedure (EOP) to the SAMG when the Core Exit Temperature (CET) reaches 923 K. Among 7 mitigation strategies envisioned, the third management strategy, injecting coolant into Reactor Coolant System (RCS) by depressurizing the RCS is one of the key strategies to fortify the in-vessel retention capability. As an aid to such mitigation strategy, a safety injection (SI) flow map has been developed by Lee et al. for predicting required flow rate associated with target RCS depressurization to recover in-vessel coolability. The mechanistic SI flow map used the CET increase rate and core water level decrease rate obtained from MELCOR code simulation. It should be noted that during the severe accident, the total heat generated is sum of the decay heat and the oxidation heat. Thus, it is necessary to consider additional heat sources by oxidation reaction between cladding, structure, control rod, and hot steam prior to the injection of coolant. It was believed that the SI model estimated required flow rate rather conservatively by considering the expected amount decay heat and the accumulated heat in the core based on the RCS pressure. However, it was confirmed that the SI model developed by Lee et al. could not represent the total heat removal especially since start of oxidation. Under transient conditions, the heat sources accumulated in the heat structures (fuel, cladding, supporting structures) should be embarked in the model for accurate contribution of the total heat source to the required flow rate. Therefore, the objective of this study is to improve the previous SI flow map predicting the injection flow rate associated with target RCS depressurization to satisfy the in-vessel coolability during severe accidents. Additional terms were developed to reflect accurate heat accumulating and included in the existing model. Resulting SI flow map includes the target RCS depressurization, which enables discharge of the required flow rate by utilizing the flow rate-pressure curve with operation of two high pressure safety injections (HPSIs). Also recent MELCOR simulation result of a postulated steam generator tube rupture (SGTR) accident verified the accuracy of the improved map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Endurance test of helium circulator for HTR-PM - 18531

High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) will be built as the world's first pebble-bed type modular HTGR commercial demonstration plant according to Chinese National science and technology program. HTR-PM will be expected to begin generating electricity in the end of 2017 in Shidao bay of China. Helium circulator is one of critical component of HTR-PM, which circulates reactor coolant through the reactor core and steam generators. The helium circulator was designed to be a vertical layout and its rotor was supported by active magnet bearings (AMBs). The whole circulator including motor and cooler was sealed in a pressure vessel. Since it is a large mechanical and electrical equipment and the design is totally advanced integrated with many new technologies, experiment study of helium circulator would be significant for safe and reliable operation of HTR-PM plant in the future. For the lack of engineering experience of this equipment, it was necessary to manufacture a prototype for tests. Meanwhile, Engineering Test Facility- Helium Circulator. ETF-HC was built in Institute of Nuclear and New Energy Technology (INET) for the tests. The prototype is completely same as the real machine for the plant. After prototype fabrication, nitrogen test was firstly conducted in the factory by the manufacturer, designer and users. After successful factory test, the prototype was delivered to INET to be tested in ETF-HC. Although nitrogen simulation test had been completed successfully, helium endurance test has been continually carried out in ETF-HC for one and half years because of several sealing problems caused by helium. After solving all kinds of problem in the field, helium circulator run continuously 50 hours under rated condition. Furthermore, 50 service cycles test was conducted in ETF-HC. More than 500 operation hours was accumulated during endurance test. Throughout the final performance test, a conclusion can be made that the helium circulator performance met the design requirements and its feasibility, availability and reliability was verified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Influence of Steam Delivery of the Working Fluid on Tool Wear

The delivery of working fluid to the cutting zone in a current of steam supplied by a steam generator is considered. This method improves the machining of metals, reduces tool wear, and improves the surface quality of the product.

42 ENGINEERING↗
DOE OSTI2018

Development of Novel Functionally Graded Transition Joints for Improving the Creep Strength of Dissimilar Metal Welds in Nuclear Applications

Steam generators of the small modular Very High Temperature Reactor (VHTR) will require use of Alloy 800 and 2.25Cr-1Mo steel in different regions depending on temperature and corrosion resistance requirements. As a result, Dissimilar Metal Welds (DMWs) between these two materials are critical for design, development, and manufacturing of the VHTR. DMWs are also used in many other applications in the power generation and petrochemical industries. Experience has demonstrated that failures of such DMWs can occur prematurely well below the expected creep life of either base metal, and these failures have significant consequences in terms of plant reliability, safety, and economic impact. These DMWs failures are caused by the abrupt change in composition that occurs at the weld interface between the two alloys. In this research, the concept of developing and optimizing Graded Transition Joints (GTJ) was developed through an integrated modeling and experimental approach. With GTJs, the composition is gradually graded between the two alloys, thus eliminating the sharp change in composition that causes failure. Microstructural modeling was first conducted to identify the optimum grade length required for improved resistance to carbon diffusion that is largely responsible for premature failure. Next, GTJs were fabricated and characterized in detail to understand how the microstructure evolves during fabrication and long term aging. Finally, the creep behavior of the GTJs was investigated and compared to that of conventional DMWs to demonstrate the superior creep strength of the GTJs. The results of this research provide significant insight into design, fabrication, microstructural evolution, and resultant performance of GTJs that can be used for reducing or eliminating DMW failures in a number of high temperature applications. The results also demonstrate that the GTJs can be readily fabricated with either conventional or additive manufacturing processes, and detailed guidelines are provided on the length of grades needed for optimal performance. The results of this research benefits the general public by improving the reliability of power generation and petrochemical plants that rely on DMWs for safe and reliable operation to deliver power and other products.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Cyber-Attack Detection and Accommodation for the Energy Delivery System

The goals of this project were to create a software system with a suite of key algorithms for cyber-attack detection and accommodation providing domain layer protection for critical power generation assets. Example assets included gas and steam turbines, heat recovery steam generators, and electrical generators. The aggressive algorithm goals were aimed at reducing the false positive rates in threat detection to <1% using learnings from many evolving disciplines (power turbine and generator physics, power system modeling, modern control theory, system identification, machine learning, deep learning, mathematics and data science). Additional goals for the algorithms involved localizing threats on-the-fly to know in which monitoring node the effects of attacks are present, and then providing accommodation to keep the system running uninterrupted much of the time in the presence of the attack. Accommodation had a performance goal of providing resiliency when up to 50% of monitoring nodes are in an attack state.

cybersecurity, cyber-physical↗
DOE OSTI2025

Synergistic heat pumped thermal storage and flexible carbon capture system

A power plant system is disclosed. The power plant system includes a combustor configured, a turbine configured to generate electricity, a heat exchanger and a steam turbine, a carbon capture system configured to remove at least a portion of carbon-based gasses from the flue gas downstream from the heat recovery steam generator, and a thermal storage system including a hot storage unit configured to store thermal energy at a hot temperature, the hot temperature greater than ambient temperature. The power plant is configured to operate in at least a first mode for storing thermal energy in the thermal storage system and a second mode for releasing the stored thermal energy from the thermal storage system and during the second mode, heat stored in the hot storage unit is transferred to the carbon capture system.

Bandhauer, Todd M.↗
DOE OSTI2015

Hydrogen-based power generation from bioethanol steam reforming

This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO{sub 2} emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

08 HYDROGEN↗
DOE OSTI2016

Plant Operation and Control Strategy for the PGSFR

The plant operating strategy and control system design for the PGSFR plant are being developed to meet functional and design requirements for the plant. The functional requirements are derived from the mission of the reactor while the design requirements are related to the reactor technologies that underlie the PGSFR. This paper describes the operation and control strategy developed to meet these requirements and the implementation of that strategy. Representative results are presented for the dynamic behavior of the nuclear steam supply system (NSSS) during operational transients. A supervisory control system has been designed for the PGSFR for Startup, Reactor Leading, and Load Following modes and it performance assessed for operational transients in the Load Following mode. The control system extends into the balance of plant to include regulation of turbine inlet pressure and steam generator pressure and includes a feedwater pump with prototypic operating characteristics. The results indicate a very stable plant response with minimal deviation from quasi-static conditions for the transients simulated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Power peaking with small modular reactors to support deep penetration of renewable energy sources - 192

As inherently intermittent sources of renewable energy more fully penetrate the US grid, peaking power is largely being supplied by carbon-emitting natural gas turbines. Significant greenhouse gas emissions can be avoided if these plants are replaced with carbon-neutral nuclear facilities to provide peaking power to complement renewable generation and meet overall power demand. There is a great deal of previous work regarding reactor power shaping with control rod movement for both currently operating nuclear power plants and proposed plants, but the literature on load-following to meet less predictable, more rapidly varying power demand is less comprehensive. We have chosen the Westinghouse IRIS integral PWR as our candidate for modeling, simulation, and control studies. A nodal model of the IRIS integral PWR, partially adapted from a MATLAB/Simulink model constructed by prior students at the University of Tennessee in Knoxville, has been constructed in Modelica. Currently the model includes the reactor vessel and steam generator system and simple assumptions for the balance of plant. The control scheme for the load-following operation of an IRIS integral PWR model would ultimately lead to the development of real operational mechanisms and principles in a grid with significant renewables capacity. In the future the model will be integrated with sophisticated grid and hybrid energy systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Preliminary Evaluation of Removing Spent Nuclear Fuel From Shutdown Sites - Fort Calhoun Site Visit - 18048

This is a technical paper that does not take into account contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. An updated evaluation of removing spent nuclear fuel (SNF) from 14 shutdown nuclear power plant sites was conducted during 2017. The major update since 2016 was the addition of the recently shutdown Fort Calhoun site to the evaluation. The Fort Calhoun site is located on the western shore of the Missouri River in Washington County in eastern Nebraska, about 31 km north of Omaha, Nebraska. Fort Calhoun has been shut down since October 24, 2016 and final removal of SNF from the reactor vessel was completed on November 13, 2016. The site visit to Fort Calhoun occurred in May 2017 and involved meetings with the Fort Calhoun staff and the Union Pacific Railroad. Participants in the site visit included the Federal Railroad Administration, Pahrump Paiute Tribe, Prairie Island Indian Community, Nebraska State Patrol, Nebraska Department of Health and Human Services, Iowa Department of Transportation Office of Rail Transportation, Iowa Department of Transportation Office of Motor Vehicle Enforcement, and the Council of State Governments-Midwest. The Fort Calhoun site was found to have two transportation mode options for the removal of SNF, rail and barge. In the case of rail, an onsite rail spur could be reinstalled from the existing Cargill industrial spur that runs through the Fort Calhoun site, or an onsite trans-load onto the Cargill industrial spur could be performed. In 2006, a barge was used to ship steam generators, a pressurizer, and a reactor vessel head to the Fort Calhoun site; consequently, barge transport of SNF offsite is potentially an option. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Systems Engineering (SE) Approach of Developing A Passive Safety Injection System (PSIS) to Cope with APR1400 Extended Station Black-Out (SBO) Accident Using RELAP5

On 11 March 2011 the power supply and cooling of the three Fukushima Daiichi reactors were disabled by a major earthquake followed by a relatively high tsunami waves. The earthquake damaged the electrical grid which is the normal power supply for the reactors, then the tsunami waves disabled the emergency diesel generators of the Fukushima Daiichi reactors, which left the plant with no alternating current (AC) source. This condition is called 'Station Black-Out'. The reactors were tripped immediately, but due to the nuclear physics phenomena, the reactor cores were still producing thermal power, which is generated by the decay of radioactive fission products in the reactor core, known as decay heat. The decay heat had not been circulated to an external heat sink and eventually the accumulative heat produced in the core led to damage to the reactor cores. After the Fukushima Daiichi accident, Design Extension Condition (DEC) such as total loss of feed water (TLOFW), and prolonged Station Black-Out (SBO) attracted the attention of nuclear experts. The attention focuses on the defense in depth concept, redundancy and sustainability of the systems, and development of the passive safety systems, which rely on natural physics laws. The topic of this paper is to discuss the possibility of attaching a Passive Safety Injection System (PSIS) to the APR1400. The PSIS is a safety system which provides adequate cooling water to the steam generators after the loss of the Auxiliary Feed Water System (AFWS) during an extended SBO accident. The PSIS will be available while the power supply of the plant is being recovered and the active safety systems take over the functionality. The anticipated PSIS will be automatically actuated to enhance the engineered safety feature and to reduce the human errors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗