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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
NASA NTRS1983

Oil Recovery System

A Downhole Steam Generation System brings oil up from deep reservoirs. The system, developed by Foster-Miller Associates consists of a steam generator, a "packer" that keeps the steam from leaking up the wellbore, and tube string that supplies air, fuel, water and hydraulics to the generator and packer; all are encased in a standard seven-inch well casing. Downhole means that the steam generator is located far down the well casing rather than on the surface. This design is more efficient than surface generated steam. A COSMIC (Computer Software Management and Information Center) program aided in the design.

Source record↗
DOE OSTI2016

Experimental study on a simplified facility of HTGR reactor building response to depressurization accidental scenarios - 18516

The analysis and understanding of air ingress events is an important aspect in the design of The High Temperature Gas Cooled Reactor (HTGR) accident scenarios including Depressurized Loss of Forced Cooling (D-LOFC) events that allow for the possibility of air ingress into the Reactor Pressure Vessel (RPV) as the result of a break in the helium pressure boundary which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. In order to characterize the air ingress into the vented low pressure compartments (VLPC) of NGNP HTGR during hypothetical small break D-LOFC accidents experimentally, a 1/28 down scaled simplified reactor building model was developed. A non-dimensional similarity approach was employed for the scaling analysis of the experimental facility. Two experiments were conducted under similar experimental conditions, with 1.1 kg/h helium injection through the top of the reactor cavity, resulting in oxygen concentrations reaching to steady state within 4 minutes. Differences between the steady state oxygen concentrations results in the steam generator cavity of the two tests with the same experimental conditions were 0.51{+-}0.086% at the top of the steam generator cavity and 0.29{+-}0.11% at the bottom of the steam generator cavity. Difference in the pressure in the reactor between the two experiments were 0.09{+-}0.026 psig. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Multi-Modular Coordination Control of HTR-PM600 Plant - 190

HTR-PM600 is a high temperature gas-cooled reactor plant based on the multi-modular scheme, i.e. six modular high temperature gas-cooled reactor (MHTGR) driving a common secondary loop system. Six MHTGR-based nuclear steam supplying system (NSSS) modules are tightly coupled by the secondary fluid flow network (FFN) which contains the secondary side of steam generators, feedwater system, main steam system, turbine generator system. The secondary FFN control, which focuses on the feedwater flowrates and main steam pressure, plays a central role in multi-modular coordination of HTR-PM600 plant. In this paper, a differential algebraic model of this secondary FFN is first proposed. Then, the corresponding flowrate-pressure control strategy is proposed, which takes the simple form of proportional-integral feedback laws. Numerical simulation results show both the feasibility and satisfactory transient performance of this controller. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

DICE-gas turbine compound reheat combined cycle

Coal-fired Direct Injection Carbon Engine – Gas Turbine (DICE-GT) Compound-Reheat Combined Cycle (CRCC), is a combined cycle power plant comprising a multiplicity of coal-fired reciprocating internal combustion engines (RICE), which is commonly referred to by the acronym DICE (Direct Injection Carbon Engine); a natural gas-fired gas turbine (turbine and combustor); a heat recovery steam generator (HRSG); a steam turbine generator; and an integrally geared and intercooled centrifugal air compressor. Finally, the concept uses coal-water slurry/fluid in the DICE and natural gas (or hydrogen) in the gas turbine for unmatched efficiency (well above 50% net LHV) and modularity (block sizes of 120, 240 and 360 MW) with easy adaptability to carbon capture and sequestration with minimal additional cost, complexity and performance hit.

42 ENGINEERING↗
DOE OSTI2013

Calculation of the dose load during dismantling of large components in the process of decommissioning of nuclear installations

The paper is focused on the decommissioning process from the point of view of steam generator dismantling. A brief description of the steam generator (used in nuclear power plants with VVER 440 type reactor) and used computer code VISIPLAN 3D ALARA are given. The main part deals with the created model and dismantling strategy together with variable input parameters - decay time and decontamination. The obtained results - external exposure of workers and the influence of time and pre-dismantling decontamination - are studied. Also detailed analyses of every dismantling step considered are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

A ranking methodology for the coupling of pressurized water nuclear reactors and molten salt thermal energy storage

Thermal energy storage (TES) systems are a proposed solution to electricity market demand fluctuations and can be coupled to nuclear power plants to enable load following. Here, this work focuses on developing a methodology for evaluating potential designs for integrating a TES system into an existing pressurized water nuclear power plant. The proposed ranking methodology allows a group of experts to postulate and weigh designs based on ranking criteria derived from literature. The methodology developed in this study can aid in the final selection of a TES design for an existing nuclear power plant. The same process could be used to analyze other TES and nuclear reactor designs. The determined optimum design from this methodology was to place the TES system after the steam generator and to utilize the steam generated from the nuclear power plant to heat a molten salt TES unit. Another conclusion of this research was that human bias is prevalent during design selection, and actions such as standardized ranking criteria and large groups of experts should be used to minimize this error.

25 ENERGY STORAGE↗
DOE OSTI2016

Experimental study on the guillotine break accident of the main steam line with Atlas

A main steam line break (MSLB) accident is defined as a pipe break in the main steam system of a nuclear power plant. The MSLB accident is characterized as cool-down event due to the increased steam flow rate, which causes excessive energy removal from the steam generators and eventually reactor coolant system. This results in a decrease in reactor coolant temperature. The cool-down causes an increase in core reactivity due to the negative moderator and Doppler reactivity coefficients. The thermalhydraulic integral effect test was performed to simulate a double-ended guillotine break accident of the main steam line of a pressurized water reactor. With an aim of simulating the MSLB accident as realistically as possible, a pertinent scaling approach was taken, especially focusing on a break flow. The main objectives of this test were not only to provide physical insight into the system response of a nuclear power plant during the MSLB accident but also to produce an integral effect test data to validate the safety analysis codes. In the present test, asymmetric thermal-hydraulic behavior was clearly observed between the loops. In the cold leg-1A and -1B where the affected steam generator is connected, the flow rates were higher and the fluid temperatures were lower than those in the cold leg-2A and -2B. This could be attributed to the excessive heat removal in the cold leg-1A and -1B through the affected steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes. Furthermore, this data can be utilized to identify any code deficiency for a MSLB simulation, especially for DVI-adapted plants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Electrification of Boilers in U.S. Manufacturing

Decarbonization of the industrial heat demand through electrification where low/no-carbon electricity is used can contribute significantly to global greenhouse gas (GHG) reduction. In U.S. manufacturing, thermal processes account for approximately 75% of the total final energy demand, of which nearly 17% was consumed by conventional industrial boilers for steam generation in 2018 (this does not include boilers for combined heat and power – CHP). Steam is generally used in industry to regulate temperatures and pressures in industrial processes, dry products, strip impurities from process fluids, etc. Although all kinds of energy sources such as fossil fuels, renewables, nuclear, and electricity can generate steam, fossil fuels’-fired boilers are dominant in U.S. manufacturing. Electric boilers, which are a mature technology, have a small market share for steam generation in the global and U.S. industry (approximately 2% in U.S. manufacturing) due to several techno-economic reasons.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2018

Coupled Calculations of the NuScale FIV Tests – Pre-test Simulations and Sensitivity Analysis

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict FIV based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the conceptual design stage. Over the past several years, the Reactor Product Line has developed a FIV capability in SHARP. SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. In previous reports we have discussed the implementation and validation of an FIV capability for helical steam generators and fuel assemblies. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo for the NuScale FIV tests. In the turbulent fretting regime one-way coupling is judged to be sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators per the low flows expected in integral PWRs. The simulations precede the tests, as experimental data was not available at the delivery of this report. Several assumptions had to be made on the support-tube interaction and multiple tests and sensitivity analysis have been performed. Additional work remains to be done and it is discussed in the future work section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Passive cooling to cold shutdown

A cooling system for a reactor module includes a reactor pressure vessel that houses primary coolant and a steam generator that lowers a temperature of the reactor pressure vessel by transferring heat from the primary coolant to a secondary coolant. The steam generator releases at least a portion of the secondary coolant as steam. Additionally, the cooling system includes a containment vessel that at least partially surrounds the reactor vessel in a containment region. The containment region is dry during normal operation of the reactor module. A controller introduces a source of water into the containment region in response to a non-emergency shut down of the reactor module. The source of water is located external to the containment vessel, and the water is introduced into the containment region after the steam generator has initially lowered the temperature of the reactor pressure vessel in response to releasing the steam.

Keller, Michael↗
DOE OSTI2016

Effects of bypass flow and power change and deviation on performance of thermal mixing structure of HTR-PM - 18424

The temperature of bypass flow not flowing through the reactor core is obviously lower than that of the main coolant helium in a High Temperature Gas-cooled Reactor (HTGR). In addition, change and deviation of reactor power because of power regulation will introduce higher radial temperature difference of coolant helium from reactor to steam generator. In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the mixing performance of the thermal mixing structure at HTR-PM reactor core outlet. The experiments on Test Facility Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation on mixing performance. In addition, some simulation calculations are conducted for the main mixing structure of the TF-HGM facility and the thermal mixing structure of HTR-PM. According to the results of experiments and simulation calculations, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, less than 30 deg. C ({+-}15 deg. C), under all considered conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Effects of wall thickness on the flow instability of supercritical pressure water in a tube - 18568

The subcritical steam generators in the current high-temperature-reactor pebble-bed module, HTR-PM, in China can be replaced by supercritical steam generators to better coordinate the reactors and the supercritical steam turbines to improve thermal efficiencies with no phase change at supercritical pressures. This study numerically studied the heat transfer and flow instability of supercritical pressure water in a vertical tube with zero- and finite-thickness walls. The effect of the wall thickness on the heat transfer of supercritical pressure water at steady state was negligible. In the transient calculations, the fluid flow rate oscillated intensely beyond a certain heat flux with zero-thickness wall, while the flows in the tube with the finite-thickness walls acted totally different. The wall properties including the density and specific heat were varied to analyze why the oscillations were suppressed when the finite-thickness wall was considered in the calculation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1986

Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system

The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

Kadambi, J. R.↗
DOE OSTI2017

Control system design for a small pressurized water reactor - 141

Small Pressurized Water Reactors (SPWR) have unique designs from the commercial large Pressurized Water Reactors (PWRs). Integral design is adopted. There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. Strong interactions exist among different loops. The pressure vessel is large in volume and the coolant inventory is also in a large amount. The inertia of the coolant is large and it takes more time for the primary system to respond to disturbances. The steam generator adopts the once-through design and the water inventory is small. The steam generator is very sensitive to disturbances. These unique characteristics challenge the control system design of SPWR. Relap5 is used to model the SPWR. Steady-state simulation is performed to verify the correctness of the model. In the reactor power control system, both the reactor fission power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance for the control system. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow and the letdown flow is compensated to calculate the charging flow needed. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

The Phenix ultimate natural convection test

The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning, a final set of tests were performed by the CEA during 9 months. Several topics were involved such as thermal hydraulics, core physics and fuel behaviour. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip of one secondary pump and a natural convection test in the primary circuit. Recognizing the unique opportunity offered by these Phenix ultimate tests, IAEA decided in 2007 to launch a Coordinated Research Project (CRP) devoted to benchmarking analyses with system codes on the Phenix natural convection test. One objective of the natural convection test in Phenix reactor is the assessment of the CATHARE system code for safety studies on future and advanced sodium cooled fast reactors. The aim of this paper is to describe this test, which was performed on June 22-23, 2009, and the associated benchmark specifications for the CRP work. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. After that, the test scenario is described: - initial state at a power of 120 MWth, - test beginning resulting from a manual dry out of the two steam generators, - manual scram, - manual trip on the three primary pumps without back-up by pony motors, - setting and development of natural convection in the primary circuit, in a first phase without significant heat sink in the secondary circuits and in a second phase with significant heat sink in the secondary circuits, by opening the casing of steam generators to create an efficient heat sink, by air natural circulation in the steam generators casing. The benchmark case ends after this second phase, which corresponds to the experimental test duration of nearly 7 hours. The paper presents also the benchmark specifications data supplied by the CEA to all participants of this CRP in order to perform calculations (core, primary circuit, primary pumps, IHX, shutdown system, operating parameters, test scenario and real test conditions). Finally, main test results and analyses are presented including the evolution of the core and of the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The natural convection has been easily set up in the pool type reactor Phenix with different boundary conditions at the secondary side, with or without heat sink. The data obtained during this unique test represent some very useful and precious results for the development of SFR in a wide range of thematic such as numerical methods dedicated to thermal-hydraulics safety analyses (system codes, CFD codes and coupling system and CFD codes) and instrumentation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Coiled Tube Gas Heaters For Nuclear Gas-Brayton Power Conversion

This project developed an alternative design for heat exchangers for application to heating supercritical carbon dioxide (S-CO 2 ) or air for power conversion. We have identified an annular coiled tube bundle configuration–where hot sodium enters tubes from multiple vertical inlet manifold pipes, flows in a spiral pattern radially inward and downward, and then exits into an equal number of vertical outlet manifold pipes–as a potentially attractive option. The S-CO 2 gas or air flows radially outward through the tube bundle. Coiled tube gas heaters (CTGHs) are expected to have excellent thermal shock, long-term thermal creep, in-service inspection, and reparability characteristics, compared to alternative options. CTGHs have significant commonality with modern nuclear steam generators. Extensive experience exists with the design, manufacture, operation, in-service inspection and maintenance of nuclear steam generators. The U.S. Nuclear Regulatory Commission also has extensive experience with regulatory guidance documented in NUREG 0800. CTGHs leverage this experience and manufacturing capability. The most important difference between steam generators and gas-Brayton cycles such as the S-CO 2 cycle is that the heat exchangers must operate with counter flow with high effectiveness to minimize the pinch-point temperature difference between the hot liquid coolant and the heated gas. S-CO 2 -cycle gas heaters also operate at sufficiently elevated temperatures that time dependent creep is important and allowable stresses are relatively low. Designing heat exchangers to operate in this regime requires configurations that minimize stresses and stress concentrations. The cylindrical tubes and cylindrical manifold pipes used in CTGHs are particularly effective geometries. The first major goal of this research project was to develop and experimentally validate a detailed, 3-D multi-phase (gas-solid-liquid) heat transport model for CTGHs, using methods similar to earlier UCB multi-scale models for PCHEs, which will enable optimization of CTGH designs with respect to the number of manifold pipes, S-CO 2 and sodium circulating power, and other design parameters. CTGHs are relatively novel because the tubes loaded in compression as well as the manifold pipes which form the tube sheets. The second major goal of the experimental and modeling effort was to optimize the assembly and heat treatment of the tube-to-tubesheet joints, using a novel tapered joint configuration, and to develop diffusion-bonded joints with high shear strength and creep resistance, as has been done in other applications for tapered plugs. The project would also conduct creep testing of these tubesheet joints in pressurized furnaces under prototypical conditions of stress and temperature, however this was not feasible due to numerous hinderances. Detailed microstructural investigations was conducted on the diffusion bond interface, while the bond strength was evaluated by tensile testing. The mechanical testing quantified the overall strength of the bond and measured the pull out force as well as long term behavior, which complemented the microstructural analysis and gave a complete understanding of the diffusion bond.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2024

Flow Instabilities in boiling channels and their suppression methodologies—A review

Small modular reactors (SMRs) are gaining significant attention as a promising solution for clean and sustainable nuclear-power generation. However, the operation of SMRs is subject to various challenges, including two-phase flow instabilities. Flow instability has the potential to trigger flow-induced vibration and cyclic fluctuations in local thermal stress. These instabilities frequently manifest because of the complex interplay among a multitude of factors, encompassing thermal-hydraulic conditions, the geometric configuration of the steam generator, and operational parameters. These conditions could subsequently lead to premature critical heat flux, equipment malfunctions, and other safety concerns. Here, the endeavor to address steam-generator flow instabilities is of utmost importance in augmenting the sustainability and efficiency of contemporary energy production. This study offers a comprehensive review of instabilities in two-phase flow, with a particular focus on the influential factors impacting the stability of flow boiling. Furthermore, it delves into the processes of identifying, characterizing, and ameliorating these instabilities, emphasizing pivotal findings, methodologies employed, and avenues for prospective research. The primary parameters of concern encompass the efficient transfer of thermal energy, the optimization of mass-flow rates, and the establishment of favorable boundary conditions, all in the context of steam generator design to alleviate instability for water-cooled SMRs. These discernments bear substantial ramifications for enhancing reactor performance and ensuring operational safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Innovative Decommissioning Strategies for Large Components - 18535

Within decommissioning projects the dismantling of steam generators as well as of the reactor pressure vessel with its internals form important and time consuming tasks. In order to optimize these, the reactor pressure vessel will be segmented into larger pieces in-situ and the larger pieces will be further segmented in a post segmentation area in order to suit the dimensional and radiological requirements of the final waste package. A similar approach is used also for the dismantling of Steam Generators. NUKEM Technologies Engineering Services has combined the two principles 'Rip / Store / Ship' and 'In-situ segmentation' to firstly segment the steam generator in-situ and then pack and ship all resulting parts allowing decontamination within 20' container to an existing decontamination facility. The aim is to free release 90 - 95% of the material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗