HTTR Demonstration Program for Nuclear Cogeneration of Hydrogen and Electricity

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Hiroyuki Sato ◽  
Xing L. Yan ◽  
Junya Sumita ◽  
Atsuhiko Terada ◽  
Yukio Tachibana
Keyword(s):  
High Temperature ◽  
Task Force ◽  
Outlet Temperature ◽  
Ministry Of Education ◽  
Reactor Outlet ◽  
Demonstration Program ◽  
Energy Agency ◽  
Sports Science ◽  
Cogeneration System ◽  
Test Reactor

The Japan Atomic Energy Agency (JAEA) initiated a high-temperature engineering test reactor (HTTR) demonstration program in accordance with recommendations of a task force established by the Ministry of Education, Culture, Sports, Science and Technology according to the Strategic Energy Plan as of April 2014. This paper explains the outline of the HTTR demonstration program with a plant concept of the heat application system directed at establishing a high-temperature gas-cooled reactor (HTGR) cogeneration system with 950°C reactor outlet temperature for production of power and hydrogen as recommended by the task force. A commercial deployment strategy including a development plan for the helium gas turbine is also presented.

Development of High Temperature Gas-Cooled Reactor Fuel for Extended Burnup

2010 ◽  
Author(s):  
Shohei Ueta ◽  
Jun Aihara ◽  
Masaki Honda ◽  
Noboru Furihata ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Coating Layer ◽  
Fission Products ◽  
Pyrolytic Carbon ◽  
Principal Function ◽  
Energy Agency ◽  
Triso Fuel ◽  
Coating Layers ◽  
Test Reactor ◽  
Fuel Particles

Current HTGRs such as the High Temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) use Tri-Isotropic (TRISO)-coated fuel particles with diameter of around 1 mm. TRISO fuel consists of a micro spherical kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and outer dense pyrolytic carbon (OPyC). The principal function of these coating layers is to retain fission products within the particle. Particularly, the SiC coating layer acts as a barrier against the diffusive release of metallic fission products and provides mechanical strength for the particle [1].

Investigation on Iodine Release Behavior During the Operation of High Temperature Engineering Test Reactor (HTTR)

2013 ◽  
Author(s):  
Shohei Ueta ◽  
Hiroyuki Inoi ◽  
Yoshitaka Mizutani ◽  
Hirofumi Ohashi ◽  
Jin Iwatsuki ◽  
...  
Keyword(s):  
High Temperature ◽  
Local Area ◽  
Release Behavior ◽  
Primary Coolant ◽  
Energy Agency ◽  
Xenon Isotopes ◽  
Test Reactor ◽  
Forced Cooling ◽  
Iodine Release ◽  
Testing Plan

Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.

Investigation of the Impact of Temperature on Hydrogen Production Cost From Advanced Water Splitting Technologies

2008 ◽  
Author(s):  
Charles O. Bolthrunis ◽  
Daniel Allen ◽  
Karl Goff ◽  
William Summers ◽  
Edward Lahoda
Keyword(s):  
High Temperature ◽  
Water Splitting ◽  
Process Temperature ◽  
Outlet Temperature ◽  
Reactor Outlet ◽  
Advantages And Disadvantages ◽  
Production Of Hydrogen ◽  
Maximum Process ◽  
Thermodynamic Limits ◽  
The Impact

One of the key technology challenges in the development of water splitting technologies is the requirement for high temperature process heat. High-Temperature Gas-Cooled Reactors (HTGRs) can supply this heat, but challenges multiply as the reactor outlet temperature, and therefore the maximum process temperature rises. A reasonable implementation strategy for applying HTGRs to these technologies would be to begin with a reactor outlet and a maximum process temperature that is achievable with today’s technology and increase those temperatures in stages as improved technology emerges. This paper investigates what those temperatures should be in the first commercial demonstration by examining the effect of these temperatures on the cost of production of hydrogen. Parameters investigated include the fundamental thermodynamic limits of each technology, reaction kinetics, materials of construction cost, process complexity, component expected life, and availability. Based on this study, comparisons are made between the leading water splitting technologies and the advantages and disadvantages of each are explained.

Sustainable Progression of Technology Education for Atomic Energy Engineering in Tsuyama National College of Technology

2013 ◽  
Author(s):  
T. Kobayashi ◽  
M. Kato ◽  
H. Sori ◽  
Y. Sasai ◽  
M. Sato ◽  
...  
Keyword(s):  
Dose Rate ◽  
Technology Education ◽  
Atomic Energy ◽  
Nuclear Reaction Analysis ◽  
Teaching Staff ◽  
Ministry Of Education ◽  
Radiation Dose Rate ◽  
Energy Agency ◽  
Sports Science ◽  
Research Activities

This study describes the achievements of a program that provides technology education about low-level radiation to develop practical core engineers. An education program starting at an early age and continuous and consistent educational agendas through seven years of college has been constructed in collaboration with regional organizations. Subjects relating to atomic energy or nuclear engineering were regrouped as “Subjects Related to Atomic Power Education” for most grades in each department. These subjects were included in the syllabus and the student guide book to emphasize a continuous and consistent policy throughout the seven-year period of college study, comprising the five-year system and the additional two-year advanced course. Furthermore, the content of lectures, experiments, and internships was enriched and realigned in collaboration with the Japan Atomic Energy Agency (JAEA), Okayama University, and Chugoku Electric Power Co., Inc. Additional educational materials were developed from inspection visits by teaching staff to atomic energy facilities were also used in the classes. Two student experiment textbooks were developed to promote two of the subjects related to atomic energy: “Cloud Chamber Experiment” and “A Test of γ-ray Inverse Square Law.” In addition to the expansion and rearrangement of atomic power education, research on atomic power conducted for graduation thesis projects was undertaken to enhance educational and research activities. Some examples are as follows: “Study on the Relation between γ Dose Rate and Rainfall in Northern Okayama Area,” “Remote Sensing of Radiation Dose Rate by Customizing an Autonomous Robot,” and “Nuclear Reaction Analysis for Composition Measurement of BN Thin Films.” It should be noted that an atomic-energy-related education working group has been in place officially to continue the above activities in the college since 2011. In consequence, although government subsidy has been decreasing, both human and material resources have been enhanced, and many students with a satisfactory understanding of atomic energy are being developed. This program was partially funded by the Ministry of Education, Culture, Sports, Science and Technology in Japan.

Modeling and Analyzing of Simple Combined Cycle of Modular High Temperature Gas Cooled Reactor

2016 ◽  
Author(s):  
Xinhe Qu ◽  
Xiaoyong Yang ◽  
Jie Wang
Keyword(s):  
High Temperature ◽  
High Efficiency ◽  
Rankine Cycle ◽  
Combined Cycle ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Brayton Cycle ◽  
Reactor Outlet ◽  
Cycle Efficiency ◽  
High Temperature Gas

High temperature gas cooled reactor (HTGR) which is one of generation IV reactor has been widely given attention in many countries since the sixties of the last century because of its inherent safety and high efficiency. Currently, the HTGR commonly uses regenerative Brayton cycle. However, as reactor outlet temperature (ROT) rising, regenerative Brayton cycle has a higher reactor inlet temperature (RIT) than 500°C and is limited by reactor materials. Combined cycle of HTGR not only can solve the problem of high RIT, but also can get a higher cycle efficiency than 50%. In this paper an accurate model of combined cycle consisting of topping Brayton cycle, bottoming Rankine cycle and heat recovery steam generator (HRSG) was established. In terms of new model of combined cycle, this paper analyzed the main properties of simple combined cycle. And put forward two optimization schemes improving the cycle efficiency of combined cycle.

Licensing Considerations of a Fluoride Salt Cooled High Temperature Test Reactor

2013 ◽  
Author(s):  
Yao Xiao ◽  
Lin-wen Hu ◽  
Charles Forsberg ◽  
Suizheng Qiu ◽  
Guanghui Su
Keyword(s):  
High Temperature ◽  
Coolant Temperature ◽  
Outlet Temperature ◽  
Fluoride Salt ◽  
Applied Physics ◽  
Reference Case ◽  
Triso Fuel ◽  
Regulatory Guidelines ◽  
Test Reactor ◽  
Temperature Test

The Fluoride salt cooled High temperature Reactor (FHR) is an innovative reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride salt cooled High temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal hydraulic licensing analysis of an FHTR using SINAP’s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting Safety System Settings (LSSS) considering uncertainties for forced convection operation are obtained. The LSSS power and coolant outlet temperature are 24.6 MW and 720 °C, respectively.

Status and Future Development for Nuclear Cogeneration System GTHTR300C

2013 ◽  
Author(s):  
Xing L. Yan ◽  
Hiroyuki Sato ◽  
Hirofumi Ohashi ◽  
Yukio Tachibana ◽  
Kazuhiko Kunitomi
Keyword(s):  
Power Generation ◽  
High Temperature ◽  
Hydrogen Production ◽  
Future Development ◽  
Heat Supply ◽  
Nuclear Reactor ◽  
Industrial Process ◽  
Outlet Temperature ◽  
Commercial Plant ◽  
Test Reactor

GTHTR300C is a small modular reactor based on a 600 MWt high temperature gas reactor (HTGR) and intended for a number of cogeneration applications such as process heat supply, hydrogen production, steelmaking, desalination in addition to power generation. The basic design has been completed by JAEA together with Japanese heavy industries. The reactor design and key plant technologies have been validated through test reactor and equipment verification. Future development includes demonstration programs to be performed on a 50 MWt system HTR50S. The demonstration programs are implemented in three steps. In the first step, a base commercial plant for heat and power is to be constructed of the same fuel proven in JAEA’s successful 950°C, 30 MWt HTGR test reactor and a conventional steam turbine such that the construction can readily proceed without major development requirement and risk. Beginning in the second step, a new fuel presently being developed at JAEA is expected to be available. With this fuel, the core outlet temperature is raised to 900°C for purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. Added in the final step is a thermochemical process to demonstrate nuclear-heated hydrogen production via water decomposition. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The designs of GTHTR300C and HTR50S will be presented and the demonstration programs will be described.

scholarly journals Performance Potential of an Advanced Nuclear Gas Turbine/Cogeneration System (HTGR-GT/C)

1983 ◽  
Author(s):  
Colin F. McDonald ◽  
Thomas H. Van Hagan ◽  
Davorin Kapich
Keyword(s):  
Gas Turbine ◽  
Waste Heat ◽  
Electrical Power ◽  
Outlet Temperature ◽  
High Grade ◽  
Closed Cycle ◽  
Performance Potential ◽  
Reactor Outlet ◽  
Cogeneration System ◽  
Steam Production

The success of the closed-cycle gas turbine depends on utilizing high-grade waste heat. This paper presents features for an advanced nuclear gas turbine operating with a reactor outlet temperature of 950°C. It discusses the results of a systems study exploring the performance potential of an advanced nuclear gas turbine/cogeneration concept (HTGR-GT/C) with a high-temperature gas-cooled reactor heat source. It also highlights the flexibility of operation with regard to combined electrical power generation and process steam production.

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