Investigation of absorption characteristics for thermal-load fluctuation using HTTR

This paper discusses the high-temperature engineering test reactor (HTTR), a 30-MW (thermal) helium gas-cooled reactor being constructed by the Japan Atomic Energy Research Establishment. A thermal mixing study of the coolant in the core bottom structure (CBS) of the HTTR is conducted to clarify the thermal-hydraulic characteristics of the coolant and estimate the influence of a hot streak on the intermediate heat exchanger (IHX) and a pressurized water cooler (PWC) down-stream from the core. An experiment is carried out using an in-core structure test section (a full-scale simulation model of the (CBS) of the helium engineering demonstration loop (HENDEL), and a numerical analysis is made using a three-dimensional time-dependent flow and heat transfer code including a k-{epsilon} model of turbulence. It is confirmed that the coolant is mixed sufficiently in the CBS and the outlet gas duct of the HTTR, and the hot streak had little effect on the IHX and the PWC.

10 years of STM

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

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 (HTGR) fuel in Japan

A Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.

Study on tritium permeation behavior in primary and second circuits of the hydrogen production system by methane steam reforming using HTGR

The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in High Temperature Gas-cooled Reactor systems.

GTHTR300 design and development

Recent advances on the use of reprocessed fuels and combined thorium fuel cycles in HTR systems

The severe-accident-free high-temperature gas-cooled reactor (SFHTR) is a prototype design for a next generation reactor. It is suitable for widespread deployment by virtue of its inherent safety features and very long refueling interval. Furthermore, its inherent safety features can be demonstrated by full-scale tests. Many of these features may be demonstrated in the High-Temperature Engineering Test Reactor (HTTR).The SFHTR is designed to have the probability of a severe accident at least two orders lower than existing systems. The fuel will not exceed its failure temperature even in the event of complete loss of coolant or complete withdrawal of two control rods. A unique configuration of burnable poisons allows a fuel cycle of 16 yr and a burnup exceeding 120 GWd/t. This feature promises very high availability and good economics.We have designed two SFHTR systems. The larger one, called the MSFHTR, has a 450- to 600-MW thermal capacity and is intended for the production of hydrogen and electricity. The smaller SFHTR (SSFHTR) is intended for remote areas, off the electrical grid, for simultaneous production of electricity and desalinated water. The SSFHTR can produce 23.5 MW(electric) plus 40 t/h of water with a net efficiency of 47%.The HTTR is capable of conducting full-scale simulation testing of key SFHTR design features in order to confirm and extend the designs and as a first step in convincing the public and the licensing authorities of the validity of demonstrable inherent safety. Design features of a 50-MW SFHTR focusing on the safety concept, safety evaluation, and core design are described. In addition, an HTTR-based test-and-development program for the SFHTR is presented.

Hydrogen produced without carbon emissions could be a useful fuel as nations look to decarbonize their electricity, transport, and industry sectors. Using the iodine–sulfur (IS) cycle coupled with a nuclear heat source is one method for producing hydrogen without the use of fossil fuels. An economic dispatch model was developed for a nuclear-driven IS system to determine hydrogen sale prices that would make such a system profitable. The system studied is the HTTR-GT/H2, a design for power and hydrogen cogeneration at the Japan Atomic Energy Agency’s High Temperature Engineering Test Reactor. This study focuses on the development of the economic model and the role that input data plays in the final calculated values. Using a historical price duration curve shows that the levelized cost of hydrogen (LCOH) or breakeven sale price of hydrogen would need to be 98.1 JPY/m3 or greater. Synthetic time histories were also used and found the LCOH to be 67.5 JPY/m3. The price duration input was found to have a significant effect on the LCOH. As such, great care should be used in these economic dispatch analyses to select reasonable input assumptions.

Cooling system design and structural integrity evaluation

The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack initiation and subcritical crack growth; field fabrication process control; property control in heavy sections; and the maintenance of high emissivity of the RPV materials over their service lifetime to enable passive heat rejection from the reactor core. All identified phenomena related to the materials of construction for the IHX, RPV, and other components were evaluated and ranked for their potential impact on reactor safety.

Application of the developed CFD analysis methodology to H2 explosion accidents in an open space