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An Analysis Of Testing Requirements For Fluoride Salt Cooled High Temperature Reactor Components
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Book Synopsis An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components by :
Download or read book An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.
Book Synopsis Fluoride-salt-cooled High-temperature Test Reactor Thermal-hydraulic Licensing and Uncertainty Propagation Analysis by : Rebecca Rose Romatoski
Download or read book Fluoride-salt-cooled High-temperature Test Reactor Thermal-hydraulic Licensing and Uncertainty Propagation Analysis written by Rebecca Rose Romatoski and published by . This book was released on 2017 with total page 307 pages. Available in PDF, EPUB and Kindle. Book excerpt: An important Fluoride-salt-cooled High-temperature Reactor (FHR) development step is to design, build, and operate a test reactor. Through a literature review, liquid-salt coolant thermophysical properties have been recommended along with their uncertainties of 2-20%. This study tackles determining the effects of these high uncertainties by proposing a newly developed methodology to incorporate uncertainty propagation in a thermal-hydraulic safety analysis for test reactor licensing. A hot channel model, Monte Carlo statistical sampling uncertainty propagation, and limiting safety systems settings (LSSS) approach are uniquely combined to ensure sufficient margin to fuel and material thermal limits during steady-state operation and to incorporate margin for high uncertainty inputs. The method calculates LSSS parameters to define safe operation. The methodology has been applied to two test reactors currently considered, the Chinese TMSR-SF1 pebble bed design and MIT's Transportable FHR prismatic core design; two candidate coolants, flibe (LiF-BeF2) and nafzirf (NaF-ZrF4); and forced flow and natural circulation conditions to compare operating regions and LSSS power (maximum power not exceeding any thermal limits). The calculated operating region accounts for uncertainty (2 [sigma]) with LSSS power (MW) for forced flow of 25.37±0.72, 22.56±1.15, 21.28±1.48, and 11.32±1.35 for pebble flibe, pebble nafzirf, prismatic flibe, and prismatic nafzirf, respectively. The pebble bed has superior heat transfer with an operating region reduced ~10% less when switching coolants and ~50% smaller uncertainty than the prismatic. The maximum fuel temperature constrains the pebble bed while the maximum coolant temperature constrains the prismatic due to different dominant heat transfer modes. Sensitivity analysis revealed 1) thermal conductivity and thus conductive heat transfer dominates in the prismatic design while convection is superior in the pebble bed, and 2) the impact of thermophysical property uncertainties are ranked in the following order: thermal conductivity, heat capacity, density, and lastly viscosity. Broadly, the methodology developed incorporates uncertainty propagation that can be used to evaluate parametric uncertainties to satisfy guidelines for non-power reactor licensing applications, and method application shows the pebble bed is more attractive for thermal-hydraulic safety. Although the method was developed and evaluated for coolant property uncertainties for FHR, it is readily applicable for any parameters of interest.
Book Synopsis High Temperature Fluoride Salt Test Loop by :
Download or read book High Temperature Fluoride Salt Test Loop written by and published by . This book was released on 2015 with total page 92 pages. Available in PDF, EPUB and Kindle. Book excerpt: Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.
Book Synopsis Heat and Mass Transfer in Packed Beds by : Noriaki Wakao
Download or read book Heat and Mass Transfer in Packed Beds written by Noriaki Wakao and published by Taylor & Francis. This book was released on 1982 with total page 396 pages. Available in PDF, EPUB and Kindle. Book excerpt: First published in 1982. Routledge is an imprint of Taylor & Francis, an informa company.
Book Synopsis The United States Fluoride Salt Cooled High Temperature Reactor Program by :
Download or read book The United States Fluoride Salt Cooled High Temperature Reactor Program written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Molten Salt Reactors and Thorium Energy by : Thomas James Dolan
Download or read book Molten Salt Reactors and Thorium Energy written by Thomas James Dolan and published by Elsevier. This book was released on 2024-01-25 with total page 1068 pages. Available in PDF, EPUB and Kindle. Book excerpt: Molten Salt Reactors and Thorium Energy, Second Edition is a fully updated comprehensive reference on the latest advances in MSR research and technology. Building on the successful first edition, Tom Dolan and the team of experts have fully updated the content to reflect the impressive advances from the last 5 years, ensuring this book continues to be the go-to reference on the topic. This new edition covers progress made in MSR design, details innovative experiments, and includes molten salt data, corrosion studies and deployment plans. The successful case studies section of the first edition have been removed, expanded, and fully updated, and are now published in a companion title called Global Case Studies on Molten Salt Reactors. Readers will gain a deep understanding of the advantages and challenges of MSR development and thorium fuel use, as well as step-by-step guidance on the latest in MSR reactor design. Each chapter provides a clear introduction, covers technical issues and includes examples and conclusions, while promoting the sustainability benefits throughout. A fully updated comprehensive handbook on Molten Salt Reactors and Thorium Energy, written by a team of global experts Covers MSR applications, technical issues, reactor types and reactor designs Includes 3 brand new chapters which reflect the latest advances in research and technology since the first edition published Presents case studies on molten salt reactors which aid in the transition to net zero by providing abundant clean, safe energy to complement wind and solar powe
Book Synopsis Fluoride-salt-cooled, High-temperature Reactor Code Benchmarking White Paper by : Edward Blandford
Download or read book Fluoride-salt-cooled, High-temperature Reactor Code Benchmarking White Paper written by Edward Blandford and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nuclear Corrosion Science and Engineering by : Damien Feron
Download or read book Nuclear Corrosion Science and Engineering written by Damien Feron and published by Elsevier. This book was released on 2012-02-21 with total page 1073 pages. Available in PDF, EPUB and Kindle. Book excerpt: Corrosion of nuclear materials, i.e. the interaction between these materials and their environments, is a major issue for plant safety as well as for operation and economic competitiveness. Understanding these corrosion mechanisms, the systems and materials they affect, and the methods to accurately measure their incidence is of critical importance to the nuclear industry. Combining assessment techniques and analytical models into this understanding allows operators to predict the service life of corrosion-affected nuclear plant materials, and to apply the most appropriate maintenance and mitigation options to ensure safe long term operation. This book critically reviews the fundamental corrosion mechanisms that affect nuclear power plants and facilities. Initial sections introduce the complex field of nuclear corrosion science, with detailed chapters on the different types of both aqueous and non aqueous corrosion mechanisms and the nuclear materials susceptible to attack from them. This is complemented by reviews of monitoring and control methodologies, as well as modelling and lifetime prediction approaches. Given that corrosion is an applied science, the final sections review corrosion issues across the range of current and next-generation nuclear reactors, and across such nuclear applications as fuel reprocessing facilities, radioactive waste storage and geological disposal systems. With its distinguished editor and international team of expert contributors, Nuclear corrosion science and engineering is an invaluable reference for nuclear metallurgists, materials scientists and engineers, as well as nuclear facility operators, regulators and consultants, and researchers and academics in this field. Comprehensively reviews the fundamental corrosion mechanisms that affect nuclear power plants and facilities Chapters assess different types of both aqueous and non aqueous corrosion mechanisms and the nuclear materials susceptible to attack from them Considers monitoring and control methodologies, as well as modelling and lifetime prediction approaches
Book Synopsis Redox Potential Measurement and Control for the Fluoride-salt-cooled High-temperature Reactor by : Kieran Patrick Dolan
Download or read book Redox Potential Measurement and Control for the Fluoride-salt-cooled High-temperature Reactor written by Kieran Patrick Dolan and published by . This book was released on 2015 with total page 93 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Optical Measurements for Fluoride-salt-cooled, High-temperature Reactors by : William Bakker Derdeyn
Download or read book Optical Measurements for Fluoride-salt-cooled, High-temperature Reactors written by William Bakker Derdeyn and published by . This book was released on 2022 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The nuclear energy industry is pursuing the commercialization of fluoride-salt-cooled, high-temperature reactors (FHRs) that operate between 500 and 700 °C, and, consequently, reactor designers need to account for radiative heat transfer (RHT) in thermal-hydraulic analysis. The absorption coefficient of the molten-fluoride coolants and emissivity of salt-exposed surfaces in the infrared range are necessary inputs for RHT modeling. The optical absorption spectroscopy techniques necessary to characterize the absorption coefficient can also be used to study the molecular structure of the liquid salt, as well as to perform in-situ measurements of the concentrations of important chemical species, such as corrosion products. In chapter 2, we report the measured emissivity of stainless steel 316 (SS316) specimens exposed to flowing FLiBe in a hot test section (HTS) at 700 °C and a cold test section (CTS) at 650 °C within the same forced-circulation loop (FCL) for 1000 hours, simulating the effect of hot and cold legs within an FHR. We conclude that the HTS coupon does not exhibit signs of air-oxidation, which means that its emissivity is likely to be representative of the in-loop surface condition and should be used for thermal-hydraulic modeling of corroded SS316 heat-exchanger surfaces in FHR designs. The reported CTS coupon emissivity could be representative of corroded heat-exchanger surfaces exposed to air after a salt-drain, but we do not recommend its use in thermal hydraulic calculations due to the presence of a thin Fe-oxide film that formed after the loop exposure. In chapter 3, we conduct a comprehensive review of methods and data from optical absorption spectroscopic measurements on molten fluorides. Transmittance, reflectance, and trans-reflectance experimental methods are discussed, along with the corresponding data reduction methodology and the limitations of each technique. Electronic and vibrational absorption data for transition-metal, lanthanide, and actinide solutes and vibrational absorption data for alkali and alkaline earth fluoride solvents are compiled, and the corresponding structural interpretation is discussed and compared with other experimental and theoretical work. We find that solvent and solute vibrational absorption can be significant in the mid-infrared, resulting in near-infrared edges of significance to RHT. Extrapolation and averaging of existing edge data leads to estimated gray absorption coefficient values at 700 °C of 546 m−1 for FLiBe and 276 m−1 for FLiNaK, both within the range of 1 - 6000 m−1 identified to be of engineering relevance for radiative heat transfer analysis. In chapter 4, we demonstrate via three different methods (peak integration, peak slope, and Beer's Law curve-fitting) the ability to correlate optical absorbance versus concentration of CrF3 and NiF2 in FLiNaK at 600 °C within the range of 100-1000 wppm. Using the correlations, we created least-squares predictive models and compared predicted concentrations with measured concentrations. For single-component solutions, the three methods all perform well giving overall average discrepancies between ±10-30 wppm. We then applied the models to multi-component solutions of NiF2 and CrF3 in FLiNaK over concentrations from 100-1000 wppm, at 600 °C. The Beer's Law curve-fitting and peak slope methods have overall average discrepancies between ±103-110 wppm, whereas the integration method has an overall average discrepancy of ±187 wppm.
Book Synopsis Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor by :
Download or read book Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.
Book Synopsis Experimental Validation of Passive Safety System Models by : Nicolas Zweibaum
Download or read book Experimental Validation of Passive Safety System Models written by Nicolas Zweibaum and published by . This book was released on 2015 with total page 231 pages. Available in PDF, EPUB and Kindle. Book excerpt: The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). The capability to validate integral transient response models is a key issue for licensing new reactor designs. This dissertation presents the scaling strategy, design and fabrication aspects, and startup testing results from the Compact Integral Effects Test (CIET) facility at UCB, which reproduces the thermal hydraulic response of an FHR under forced and natural circulation operation. CIET provides validation data to confirm the performance of the direct reactor auxiliary cooling system (DRACS) in an FHR, used for natural-circulation-driven decay heat removal, under a set of reference licensing basis events, as predicted by best-estimate codes such as RELAP5-3D. CIET uses a simulant fluid, Dowtherm A oil, which at relatively low temperatures (50-120°C) matches the Prandtl, Reynolds, Froude and Grashof numbers of the major liquid salts simultaneously, at approximately 50% geometric scale and heater power under 2% of prototypical conditions. The studies reported here include isothermal pressure drop tests performed during startup testing of CIET, with extensive pressure data collection to determine friction losses in the system, as well as subsequent heated tests, from parasitic heat loss tests to more complex feedback control tests and natural circulation experiments. For initial code validation, coupled steady-state single-phase natural circulation loops and simple forced cooling transients were conducted in CIET. For various heat input levels and temperature boundary conditions, fluid mass flow rates and temperatures were compared between RELAP5- 3D results, analytical solutions when available, and experimental data. This study shows that RELAP5-3D provides excellent predictions of steady-state natural circulation and simple transient forced cooling in CIET. The code predicts natural circulation mass flow rates within 8%, and steady-state and transient fluid temperatures, under both natural and forced circulation, within 2°C of experimental data, suggesting that RELAP5-3D is a good EM to use to design and license FHRs. A key element in design and licensing of new reactor technology lies in the analysis of the plant response to a variety of potential transients. When applicable, this involves understanding of passive safety system behavior. This dissertation develops a framework to assess reliability and propose design optimization and risk mitigation strategies associated with passive decay heat removal systems, applied to the Mk1 PB-FHR DRACS. This investigation builds upon previous detailed design work for Mk1 components and the use of RELAP5-3D models validated for FHR natural circulation phenomenology. For risk assessment, reliability of the point design of the passive safety system for the Mk1 PB-FHR, which depends on the ability of various structures to fulfill their safety functions, is studied. Whereas traditional probabilistic risk assessment (PRA) methods are based on event and fault trees for components of the system that perform in a binary way - operating or not operating -, this study is mostly based on probability distributions of heat load compared to the capacity of the system to remove heat, as recommended by the reliability methods for passive safety functions (RMPS) that are used here. To reduce computational time, the use of response surfaces to describe the system in a simplified manner, in the context of RMPS, is also demonstrated. The design optimization and risk mitigation part proposes a framework to study the elements of the design of the reactor, and more specifically its passive safety cooling system, which can contribute to enhanced reliability of heat removal under accident conditions. Risk mitigation measures based on design, startup testing, in-service inspection and online monitoring are proposed to narrow probability distributions of key parameters of the system and increase reliability and safety. Another major aspect in the development of novel energy systems is the assessment of their impacts on the environment compared to current technologies. While most existing life cycle assessment (LCA) studies have been applied to conventional nuclear power plants, this dissertation proposes a framework to extend such studies to advanced reactor designs, using the example of the Mk1 PB-FHR. The Mk1 uses a nuclear air-Brayton combined cycle designed to produce 100 MWe of base-load electricity when operated with only nuclear heat, and 242 MWe using natural gas co-firing for peaking power. The Mk1 design provides a basis for quantities and costs of major classes of materials involved in building the reactor and fabricating fuel, and operation parameters. Existing data and economic input-output LCA models are used to calculate greenhouse gas emissions per kWh of electricity produced over the life cycle of the reactor. Baseline life cycle emissions from the Mk1 PB-FHR in base-load configuration are 26% lower than average Generation II light water reactors in the U.S., 98% lower than average U.S. coal plants and 96% lower than average U.S. natural gas combined cycle plants using the same turbine technology. In peaking configuration, due to its nuclear component and higher thermal efficiency, the Mk1 plant only produces 32% of the emissions of average U.S. gas turbine simple cycle peaking plants. One key contribution to life cycle emissions results from the amount and type of concrete used for reactor construction. This is an incentive to develop innovative construction methods using optimized steel-concrete composite wall modules and new concrete mixes to reduce life cycle emissions from the Mk1 and other advanced reactor designs.
Book Synopsis Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment by : Jyeshtharaj Joshi
Download or read book Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment written by Jyeshtharaj Joshi and published by Woodhead Publishing. This book was released on 2019-06-09 with total page 888 pages. Available in PDF, EPUB and Kindle. Book excerpt: Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment presents the latest computational fluid dynamic technologies. It includes an evaluation of safety systems for reactors using CFD and their design, the modeling of Severe Accident Phenomena Using CFD, Model Development for Two-phase Flows, and Applications for Sodium and Molten Salt Reactor Designs. Editors Joshi and Nayak have an invaluable wealth of experience that enables them to comment on the development of CFD models, the technologies currently in practice, and the future of CFD in nuclear reactors. Readers will find a thematic discussion on each aspect of CFD applications for the design and safety assessment of Gen II to Gen IV reactor concepts that will help them develop cost reduction strategies for nuclear power plants. Presents a thematic and comprehensive discussion on each aspect of CFD applications for the design and safety assessment of nuclear reactors Provides an historical review of the development of CFD models, discusses state-of-the-art concepts, and takes an applied and analytic look toward the future Includes CFD tools and simulations to advise and guide the reader through enhancing cost effectiveness, safety and performance optimization
Book Synopsis Molten-Salt Reactor Program Semiannual Progress Report for Period Ending ... by : Oak Ridge National Laboratory
Download or read book Molten-Salt Reactor Program Semiannual Progress Report for Period Ending ... written by Oak Ridge National Laboratory and published by . This book was released on 1965-02 with total page 198 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Electrochemical Cell Design by : R.E. White
Download or read book Electrochemical Cell Design written by R.E. White and published by Springer Science & Business Media. This book was released on 2012-12-06 with total page 397 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Buoyancy-induced Flows and Transport by : Benjamin Gebhart
Download or read book Buoyancy-induced Flows and Transport written by Benjamin Gebhart and published by CRC Press. This book was released on 1988 with total page 594 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nuclear Energy for Hydrogen Generation through Intermediate Heat Exchangers by : Bahman Zohuri
Download or read book Nuclear Energy for Hydrogen Generation through Intermediate Heat Exchangers written by Bahman Zohuri and published by Springer. This book was released on 2016-07-15 with total page 421 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book describes recent technological developments in next generation nuclear reactors that have created renewed interest in nuclear process heat for industrial applications. The author’s discussion mirrors the industry’s emerging focus on combined cycle Next Generation Nuclear Plants’ (NGNP) seemingly natural fit in producing electricity and process heat for hydrogen production. To utilize this process heat, engineers must uncover a thermal device that can transfer the thermal energy from the NGNP to the hydrogen plant in the most performance efficient and cost effective way possible. This book is written around that vital quest, and the author describes the usefulness of the Intermediate Heat Exchanger (IHX) as a possible solution. The option to transfer heat and thermal energy via a single-phase forced convection loop where fluid is mechanically pumped between the heat exchangers at the nuclear and hydrogen plants is presented, and challenges associated with this tactic are discussed. As a second option, heat pipes and thermosyphons, with their ability to transport very large quantities of heat over relatively long distance with small temperature losses, are also examined.
