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Author: Nihon Genshiryoku Kenkyūjo. Division of High Temperature Engineering
Publisher:
ISBN:
Category : Gas cooled reactors
Languages : en
Pages : 43
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Book Description
Author: Nihon Genshiryoku Kenkyūjo. Division of High Temperature Engineering
Publisher:
ISBN:
Category : Gas cooled reactors
Languages : en
Pages : 43
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Book Description
Author: Tetsuaki Takeda
Publisher: Academic Press
ISBN: 012821032X
Category : Business & Economics
Languages : en
Pages : 478
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Book Description
High-Temperature Gas Reactors is the fifth volume in the JSME Series on Thermal and Nuclear Power Generation. Series Editor Yasuo Koizumi and his Volume editors Tetsuaki Takeda and Yoshiyuki Inagaki present the latest research on High-Temperature Gas Reactor (HTGR) development and utilization, beginning with an analysis of the history of HTGRs. A detailed analysis of HTGR design features, including reactor core design, cooling tower design, pressure vessel design, I&C factors and safety design, provides readers with a solid understanding of how to develop efficient and safe HTGR within a nuclear power plant. The authors combine their knowledge to present a guide on the safety of HTGRs throughout the entire reactor system, drawing on their unique experience to pass on lessons learned and best practices to support professionals and researchers in their design and operation of these advanced reactor types. Case studies of critical testing carried out by the authors provide the reader with firsthand information on how to conduct tests safely and effectively and an understanding of which responses are required in unexpected incidents to achieve their research objectives. An analysis of technologies and systems in development and testing stages offer the reader a look to the future of HTGRs and help to direct and inform their further research in heat transfer, fluid-dynamics, fuel options and advanced reactor facility selection. This volume is of interest for nuclear and thermal energy engineers and researchers focusing on HTGRs, HTGR plant designers and operators, regulators, post graduate students of nuclear engineering, national labs, government officials and agencies in power and energy policy and regulations. Written by the leaders and pioneers in nuclear research at the Japanese Society of Mechanical Engineers and draws upon their combined wealth of knowledge and experience Includes real examples and case studies from Japan, the US and Europe to provide a deeper learning opportunity with practical benefits Considers the societal impact and sustainability concerns and goals throughout the discussion Includes safety factors and considerations, as well as unique results from performance testing of HTGR systems.
Author: Kurt Kugeler
Publisher: Springer
ISBN: 3662577127
Category : Technology & Engineering
Languages : en
Pages : 903
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Book Description
"Modular High-temperature Gas-cooled Reactor Power Plant" introduces the power plants driven by modular high temperature gas-cooled reactors (HTR), which are characterized by their inherent safety features and high output temperatures. HTRs have the potential to be adopted near demand side to supply both electricity and process heat, directly replacing conventional fossil fuels. The world is confronted with two dilemmas in the energy sector, namely climate change and energy supply security. HTRs have the potential to significantly alleviate these concerns. This book will provide readers with a thorough understanding of HTRs, their history, principles, and fields of application. The book is intended for researchers and engineers involved with nuclear engineering and energy technology.
Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201253101
Category : Business & Economics
Languages : en
Pages : 639
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Book Description
This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 51
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Book Description
The Modular High Temperature Gas-Cooled Reactor (MHTGR) currently under development by the US Department of Energy (US-DOE) for commercial applications has top-level goals of producing safe, economical power for the US utility industry. The utility industry has been represented in formulating design and licensing requirements through both a Utility User Requirements Document'' and by participating in the DOE system engineering process known as the Integrated Approach.'' The result of this collaboration has been to set stringent goals for both the safety and operational reliability of the MHTGR. To achieve these goals, the designer must have access to a more comprehensive data base of properties in several fields of technology than is currently available. A technology development program has been planned to provide this data to the designer in time to support both his design activities and the submittal of formal licensing application documents. The US-DOE has chosen the Oak Ridge National Laboratory (ORNL) to take the lead in planning and executing these technology programs. When completed these will augment the designer's current data base and provide the necessary depth to meet the stringent goals which have been set for the MHTGR. It is worth noting that the goals of safety and operational reliability are complementary, and the data required from the technology development program will be similar. Therefore, the program to support the licensing of the MHTGR is not separate from that required for design, but is a subset of that which meets all the requirements that result from implementing the US-DOE's integrated approach. 38 figs.
Author:
Publisher:
ISBN:
Category : Ocean thermal power plants
Languages : en
Pages : 70
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Book Description
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R & D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R & D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.
Author: U.S. Atomic Energy Commission. Division of Reactor Development and Technology
Publisher:
ISBN:
Category : Gas cooled reactors
Languages : en
Pages : 56
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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133 Mw-hr of fission heat, was disassembled and examined. No fuel-compact damage of any type was visible. Determination of the distribution of fission products in the element is under way, Fissionproduct- release data for in-pile-loop element III-A were calculated. During the 133 Mw- hr of operation, the release fraction increased by approximately one order of magnitude. Also calculated were the xenon and krypton release data for the first 100 Mw-hr of III-B operation. The release rate for the longer-lived isotopes increased bv about a factor of 10 and that of the shorter-lived isotopes by about a factor of 100. A test was run in which the in-pileloop purge flow, was stopped. The primariy-loop activity level rose sharply during the first hour, increased at a slower rate for the next 11 hr, and then appeared to level off. When purge flow was resumed, the gross activity in the primary loop was cleaned up with a half life of about 2.2 hr. An attempt was made to identify Cs/sup 137/ and Ba/ sup 140/ plateout in portions of the in-pile loop. A very small amount of cesium (less than a monolayer) was found, but no barium could be detected. The validity of two basic assumptions made in the one-dimensional burnup code FEVER was investigated. As a result of extensive lifetime studies and power-distribution and temperaturecoefficient calculations, the initial fuel loading for the Peach Bottom core was specified. A series of control-rodworth calculations and a recalculation of the postulated rod-fall accident were made for this loading. The test of the prototype control rod and drive was satisfactorily coNonempleted during the quarter. During the course of the test the drive completed 590,641 starts and stops, 5,756 scrams, and more than 2.6 million inches of random regulating motion in helium at reactor temperatures. These totals far exceed the expected life requirements of the system. Preparations are being made for testing the prototype emergency shutdown rod and drive. The apparatus for the barium permeation experiment with a full-diameter sleeve was completed, and preliminary calibration runs were started. Following these runs, the system will be operated until an equilibrium distribution of barium is reached. At that time, a series of corings will be made on all of the compacts and the sleeve to evaluate the overall barium and strontium distribution. Other experiments on barium behavior, including permeation experiments with reducedscale fuel elements and exVeriments on the vaporization, sorption, and diffusion of barium, were continued, and the data are being analyzed. Measurements were made to compare the room-temperature back diffusion of argon, krypton, and xenon through a sample of sleeve graphite against a helium pressure difference. The results show that the difference between the effective back-diffusion coefficients of krypton and xenon seems to increase with increasing helium pressure difference across the sleeve. The argon and krypton back-diffusion data at an average pressure of 3 atm are essentially the same, A FORTRAN code was written to recalculate the retention of neutron poison material and fission products in the core as well as their condensation on and revaporization from the upper reflector following a complete loss-of-coolant-circulation accident. (auth).
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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The HTGR prototype plant (Peach Bottom Power Reactor) is being designed to produce steam at l450 psi and 1000 deg F and to have a net capacity of 40 Mw(e). The fuel temperatures and gas pressures will be approximately the same as those required for larger plants. The reactor data and operating conditions for the graphite-clad core are given. The reactor and primary coolant systems are described. The prospects for development of the graphite-clad fuel element in time for use in the first loading of the reactor were improved by important advances in methods of fabrication and testing of both fuel compacts and graphite sleeves. The hot-pressing process for making fuel compacts was used successfully to make full-size compacts with a uniform distribution of ThC2- UC2 particles. Three irradiation capsules were fabricated and inserted in a test reactor to determine fuel compact and sleeve performance under HTGR conditions of irradiation and temperature. Two of these ran satisfactorily for the scheduled time of operation. A scope design study of the in-pile loop that will be used to evaluate the full-diameter graphite-clad element was completed. Experiments to determine the extent of fuel migration within the element were undertaken. Preliminary results indicated that the central fuel-element temperatures must not exceed 2300-C for routine operation. An important start was made in developing an understanding of how to treat the neutron thermalization process in high-temperature graphite reactors. Analytical techniques for calculating the thermal neutron spectra in poisoned graphite media were developed and programmed for the IBM 704 computer. The experimental technique of measuring neutron spectra by using a pulsed linear electron accelerator was demonstrated by measurements made with boron-loaded graphite. A mockup of a small portion of the reactor core was constructed and operated to determine the local heat-transfer coefficients and pressure drop in the tricusp- shaped coolant passage. Initial results indicated that the variation of the heat-transfer coefficient around the circumference of the element is less than expected. Studies were started of the transient temperatures and stresses developed in the pressure vessel as a result of load changes or a scram. A detailed study of several types of steam generator for use in the nuclear steam supply system was completed. A design incorporating a steam drum was selected for further study. Preliminary flow diagrams were completed for the helium- purification and fission-product trapping systems. Adsorption isobars for selected fission products in activated carbon were measured and will be used in the detailed design of the trapping system. Detailed planning of the experimental reactor physics program was initiated. Progress was made in the identification of the principal safeguards problems for this type of reactor, and a preliminary safety analysis of the plant was completed. (auth).
