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At least 109 records · Page 6

All sources NASA NTRS DOE OSTI

DOE OSTI2017

TRACE Analysis of Loss of Flow for the APR1400

The APR1400 is a pressurized water reactor (PWR). The APR1400 nuclear steam supply system consists of a reactor pressure vessel (RPV), which contains the reactor core, and two primary reactor coolant system loops connected in parallel to the RPV. Each coolant loop consists of a hot leg (HL), a u-tube steam generator (SG), two reactor coolant pumps (RCPs) and two cold legs (CLs). A pressurizer is connected to the hot leg on one of the coolant loops. During normal operation, the RCPs circulate high pressure water through the reactor core to remove the heat. The heated water exits the RPV and passes through the hot legs into the steam generator tubes. Heat is transferred through the walls of the steam generator tubes into a lower pressure secondary coolant system. Cooled primary system water from the steam generators is returned through the RCPs and cold legs to the RPV. The RCPs operate at constant speed; the primary coolant system is designed to provide a prescribed rate of coolant flow through the reactor core. The APR1400 SG secondary coolant system pressure is controlled through modulation of valves in the steam lines to maintain a prescribed turbine inlet pressure. The SG secondary coolant inventory is controlled by using a feedwater control system to maintain a prescribed liquid level in the steam generator downcomer. This water inventory level is sufficient to cover the outer surfaces of the steam generator tubes while preventing overfill that would cause liquid to pass from the steam generators through the steam lines into the turbine system. In the current work, the Nuclear Regulatory Commission's reactor systems analysis code, TRACE?2 was used to perform an analysis of a loss of flow anticipated operational occurrence (AOO). The loss of flow is initiated by a loss of off-site power (LOOP), resulting in a trip of the RCPs. The regulatory purpose of the current work is to confirm safety analyses documented in the APR1400 Design Control Document (DCD)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Simulated performance of biomass gasification based combined power and refrigeration plant for community scale application

Thermal performance analysis and sizing of a biomass gasification based combined power and refrigeration plant (CPR) is reported in this study. The plant is capable of producing 100 kWe of electrical output while simultaneously producing a refrigeration effect, varying from 28-68 ton of refrigeration (TR). The topping gas turbine cycle is an indirectly heated all-air cycle. A combustor heat exchanger duplex (CHX) unit burns producer gas and transfer heat to air. This arrangement avoids complex gas cleaning requirements for the biomass-derived producer gas. The exhaust air of the topping GT is utilized to run a bottoming ammonia absorption refrigeration (AAR) cycle via a heat recovery steam generator (HRSG), steam produced in the HRSG supplying heat to the generator of the refrigeration cycle. Effects of major operating parameters like topping cycle pressure ratio (r{sub p}) and turbine inlet temperature (TIT) on the energetic performance of the plant are studied. Energetic performance of the plant is evaluated via energy efficiency, required biomass consumption and fuel energy savings ratio (FESR). The FESR calculation method is significant for indicating the savings in fuel of a combined power and process heat plant instead of separate plants for power and process heat. The study reveals that, topping cycle attains maximum power efficiency of 30%in pressure ratio range of 8-10. Up to a certain value of pressure ratio the required air flow rate through the GT unit decreases with increase in pressure ratio and then increases with further increase in pressure ratio. The capacity of refrigeration of the AAR unit initially decreases up to a certain value of topping GT cycle pressure ratio and then increases with further increase in pressure ratio. The FESR is found to be maximized at a pressure ratio of 9 (when TIT=1100°C), the maximum value being 53%. The FESR is higher for higher TIT. The heat exchanger sizing is also influenced by the topping cycle pressure ratio and GT-TIT.

09 BIOMASS FUELS↗

DOE OSTI2015

Waste remediation

A system including a steam generation system and a chamber. The steam generation system includes a complex and the steam generation system is configured to receive water, concentrate electromagnetic (EM) radiation received from an EM radiation source, apply the EM radiation to the complex, where the complex absorbs the EM radiation to generate heat, and transform, using the heat generated by the complex, the water to steam. The chamber is configured to receive the steam and an object, wherein the object is of medical waste, medical equipment, fabric, and fecal matter.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2017

Waste remediation

60 APPLIED LIFE SCIENCES↗

NASA NTRS1981

The small community solar thermal power experiment. Parabolic dish technology for industrial process heat application

Aspects of incorporating a thermal energy transport system (ETS) into a field of parabolic dish collectors for industrial process heat (IPH) applications were investigated. Specific objectives are to: (1) verify the mathematical optimization of pipe diameters and insulation thicknesses calculated by a computer code; (2) verify the cost model for pipe network costs using conventional pipe network construction; (3) develop a design and the associated production costs for incorporating risers and downcomers on a low cost concentrator (LCC); (4) investigate the cost reduction of using unconventional pipe construction technology. The pipe network design and costs for a particular IPH application, specifically solar thermally enhanced oil recovery (STEOR) are analyzed. The application involves the hybrid operation of a solar powered steam generator in conjunction with a steam generator using fossil fuels to generate STEOR steam for wells. It is concluded that the STEOR application provides a baseline pipe network geometry used for optimization studies of pipe diameter and insulation thickness, and for development of comparative cost data, and operating parameters for the design of riser/downcomer modifications to the low cost concentrator.

Polzien, R. E.↗

NASA NTRS1981

Steam Rankine Solar Receiver, phase 2

A steam rankine solar receiver (SRSR) based on a tubular concept was designed and developed. The SRSR is an insulated, cylindrical coiled tube boiler which is mounted at the focal plane of a fully tracking parabolic solar reflector. The concentrated solar energy received at the focal plane is then transformed to thermal energy through steam generation. The steam is used in a small Rankine cycle heat engine to drive a generator for the production of electrical energy. The SRSR was designed to have a dual mode capability, performing as a once through boiler with and without reheat. This was achieved by means of two coils which constitute the boiler. The boiler core size of the SRSR is 17.0 inches in diameter and 21.5 inches long. The tube size is 7/16 inch I.D. by 0.070 inch wall for the primary, and 3/4 inch I.D. by 0.125 inch wall for the reheat section. The materials used were corrosion resistant steel (CRES) type 321 and type 347 stainless steel. The core is insulated with 6 inches of cerablanket insulation wrapped around the outer wall. The aperture end and the reflector back plate at the closed end section are made of silicon carbide. The SRSR accepts 85 kwth and has a design life of 10,000 hrs when producing steam at 1400 F and 2550 psig.

Deanda, L. E.↗

DOE OSTI2019

Collaboration for the Management of the Bruce Power Major Component Replacement (MCR) Nuclear Waste - 19630

Bruce Power's Major Component Replacement (MCR) Program will generate large volume waste via steam generators (SGs) and intermediate-level nuclear waste, which will require specialized packaging and handling to safely receive and store it at Ontario Power Generation's (OPG) Western Waste Management Facility (WWMF). The refurbishment of a Candu reactor involves, in part, the removal and replacement of all of a reactor core's fuel channel assemblies. Bruce Power and OPG collaborated to develop nuclear waste management solutions to optimize the use of OPG's existing Steam Generator Storage Building (SGSB) and Retube Component Storage Building (RCSB). By designing specialized Retube Waste Containers (RWCs) and optimizing configurations for the waste to be stored in the buildings, we will reduce the interim nuclear waste storage needs at the OPG WWMF, while improving Bruce Power MCR process efficiencies and reducing the overall costs of the MCR Program (Units 3 through 8). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

CRUD Source Term Assessment and Development for VERA

In pressurized water reactors, corrosion of primary loop materials is an ongoing process, as primary loop components (steam generator, hot leg, cold leg, and core) are constantly exposed to the coolant. As a result of this corrosion, primary loop materials are released into the coolant in the form of aqueous metal ions and particulate oxides. These corrosion products are the source of materials that form crud deposits on fuel rod clad surfaces. The mass of corrosion products available to create crud will limit the rate and extent of crud buildup that occurs. This is commonly referred to as the crud “source term”. The major contribution to the source term mass comes from the steam generator, due to the large surface area exposed to the coolant. Steam generators are not identical; the wetted area exposed to the coolant and the alloy used in the steam generator tubes vary. Different alloys corrode at different rates; Inconel 690 corrodes at approximately one-third of the rate of corrosion of Inconel 600. The mass of corrosion products released from the steam generator also varies by the area exposed to the coolant. The release of these corrosion products depends on the local environmental factors within the primary loop, including temperature, pH, local fluid conditions, electrochemical potential, and thermodynamic solubilities of the species present within this system. As such, each PWR will have a different crud source term based on these plant-specific parameters. In order to accurately model and predict crud growth and issues that arise from the presence of crud (CIPS, CILC), the source of these corrosion product materials must be accurately modeled in a system mass balance. These models should be mechanistic, representative of the underlying physics phenomena, and account for the effects of the plant-specific design and operating parameters, including the alloys used, local temperatures, and fluid flow conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Development of Scaling Criteria for Modeling Condensation in Horizontal Tubes Using Reduced Pressure System and Simulant Fluid

Most of the currently operating nuclear power plants (NPPs), generation-II and III, are equipped with active safety systems, which raises concerns regarding those safety systems' operational reliability and the suitability of the accidents prevention measures. Following the severe accidents at nuclear power plants at Three Mile Island, Chernobyl, and recently in Fukushima some advance Pressurized Water Reactors (PWRs) have adopted passively cooled steam generator natural circulation system to remove the decay heat passively through the steam generator (SG). This is done by condensing steam from the steam generator inside a heat exchanger submerged in a tank of water or an air cooled system. The APR+ (Advanced Power Reactor Plus), which is currently being developed in South Korea, has adopted a Passive Auxiliary Feedwater System (PAFS), which removes the decay heat through the steam generator replacing the conventional active auxiliary feedwater system. The main part of the PAFS, in where the steam condenses, is the Passive Condensation Heat Exchanger (PCHX). The steam condenses under high pressure (7.4 MPa, 290 deg. C) in 8.4 m long horizontal tube, with 3 degrees inclination, with inner diameter of 44.8 mm and wall thickness of 3 mm. The PCHX consists of 4 bundles, each bundle has 60 condensation tubes. The performance of the PCHX performance should be evaluated. Condensation has been extensively studied; however, most previous experimental work was directed to low pressure (i.e., near atmospheric pressure) and small condensation tubes. Moreover, studies on the horizontal type heat exchanger are limited, since previous safety systems and experimentation work were directed to the vertical type heat exchangers. Due to the high construction cost of a full size high pressure condensation experiment, a reduced-size reduced-pressure condensation experiment is suggested to evaluate the heat removal performance of a condensation tube of the PAFS heat exchanger. A scaling methodology to be established to design a reduced-size reduced-pressure condensation experiment. This study attempts to develop a scaling criteria for the PAFS condensation system, and to apply it for modeling a high pressure water system by a low pressure Freon system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Degradation Mechanisms and Development of Protective Coatings for Thermal Energy Storage and High Temperature Fluid Containment Materials

This research addresses cost savings to achieve the SETO target of $15/kWh th for thermal energy storage and 6¢/kWh for levelized cost of energy. Next-generation solar power conversion systems require high-temperature, advanced fluids in the range of 600° to 900°C. The three major advanced fluid groups—supercritical CO 2 , liquid metal alloys, and molten salts—are very corrosive to common alloys used in vessels, heat exchangers, and piping at these temperatures. Understanding degradation mechanisms is essential for making recommendations on proper working conditions and suitable cost-effective coating materials for CSP applications at high temperatures. The implementation of the findings from this research will significantly reduce investment risk by increasing the lifetime of the CSP plant components. In the past, the majority of the hot corrosion evaluations have been performed for waste incinerators, gas turbine engines, and electric power generation (steam-generating equipment) applications for different materials, molten-salt systems, and coatings. Containment materials and resistant coatings for supercritical CO 2 , molten salts and liquid metal alloys have not being evaluated for CSP applications. We plan to understand the underlying degradation mechanisms to select protective approaches such as anodic passivation, cathodic protection, and coating designs.The novelty of the proposed research is the development of protective coatings and working conditions for materials in contact with high temperature TES and HTF. Current state-of-the-art materials and protection methods are non-existent. Novel approaches are: the use of graphite nanomaterials to enhance the non-wettability of the aggressive fluids with the coating; self-healing behavior of Si-coatings in liquid Al-Si alloys; anodic passivation in molten salts to protect metallic substrates and the modeling prediction of attack in liquid metal alloys.

14 SOLAR ENERGY↗

DOE OSTI2023

Dynamic Modeling of a Solar-To-Hydrogen Flexible High Temperature Steam Electrolysis Plant

Sustainble hydrogen production for use as a renewable combustible fuel and clean chemical feedstock is an important objective as the world moves towards a renewable energy future. High temperature steam electrolysis is a promising hydrogen production technology due to its reduced electric input that is offset by heat input into steam generation and steam superheating. An option to provide this heat is to use concentrating solar thermal technology that can sustainably provide heat input while renewable electricity is used for the electrolysis reaction. In this work, a solar-to-hydrogen high temperature steam electrolysis plant is designed and dynamically modeled, showing continuous hydrogen production by utilizing supplemental heating and efficient recuperative heating from the electrolysis product streams. Through this design, over 90% of the required heat input for the process can by met by a combination of solar and recuperative heat. Additionally, the plant can flexibility operate by ramping down hydrogen production and through flexible heat integration, which intelligently integrates solar heat based on solar conditions. Smooth operation with flexible hydrogen production is demonstrated which decreases electrical input during on-peak grid times and also decreases the total supplemental heat load over the course of a day from 26.1% to 24.5%. In addition, by using flexible heat integration, the plant can increase its solar heat usage by 4.1% relative to a base case. Both options for flexibility show efficient use of solar thermal energy to sustainably and continuously produce hydrogen.

Immonen, Jake (ORCID:0000000341231625)↗

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 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↗