Masuro Ogawa

Experiments on heat transfer of smooth and swirl tubes under one-sided heating conditions

To design the divertor plate for the next generation fusion machines which is subjected to high heat loads on its one side by the plasma, it is essential to evaluate performances of heat transfer efficiency. However there is little in the literature for predicting of heat transfer coefficients under such one-sided heating conditions. To establish the heat transfer correlation for water under one-sided heating conditions, the authors have performed heat transfer experiments on smooth circular and swirl tubes in the regions from non-boiling to high subcooled partial nucleate boiling. Based on the experimental results, it is confirmed that the existing heat transfer correlations can be applicable at the non-boiling region. For the subcooled partial nucleate boiling region, they cannot be available so that a new heat transfer correlation has been proposed under one-sided heating conditions.

Burnout experiments on the externally-finned swirl tube for steady-state and high-heat flux beam stops

An experimental study to develop beam stops for the next generation of neutral beam injectors was started, using an ion source developed for the JT-60 neutral beam injector. A swirl tube is one of the most promising candidates for a beam stop element which can handle steady-state and high-heat flux beams. In the present experiments, a modified swirl tube, namely an externally-finned swirl tube, was tested together with a simple smooth tube, an externally finned tube, and an internally finned tube. The major dimensions of the tubes are 10 mm in outer-diameter, 1.5 mm in wall thickness, 15 mm in external fin width, and 700 mm in length. The burnout heat flux (CHF) normal to the externally finned swirl tube was 4.1 ± 0.1 kW/cm2, where the Gaussian e-folding half-width of the beam intensity distribution was about 90 mm, the flow rate of the cooling water was 30 l/min, inlet and outlet gauge pressures were about 1 MPa and 0.2 MPa, respectively, and the temperature of the inlet water was kept to 20 °C during a pulse. A burnout heat flux ratio, which is defined by the ratio of the CHF value of the externally-finned swirl tube to that of the externally-finned tube, turned out to be about 1.5. Burnout heat fluxes of the tubes with a swirl tape or internal fins increase linearly with an increase of the flow rate. It was found that the tube with external fins has effects that not only reduce the thermal stress but also improve the characteristics of boiling heat transfer.

Thermal shock tests on various materials of plasma facing components for FER/ITER

Development of plasma facing components and materials is a key element in the R&D program for the Fusion Experimental Reactor (FER), which has been designed at JAERI, and the International Thermonuclear Experimental Reactor (ITER), which has been designed under international collaboration. In these next-step tokamak devices, the plasma facing components and materials will be exposed to severe heat load and incident particle flux. The concern is especially acute that the extremely high thermal shock due to plasma disruption could cause material fracture. Efforts on developing the first wall and divertor have been energetically undertaken at JAERI. The present paper describes recent experimental and analytical results on thermal shock characteristics of various materials.

Experimental study on melting and evaporation of metal exposed to intense hydrogen ion beam

Abstract This report describes experimental studies on melting and evaporation of metals, mainly stainless steels, subjected to high heat flux simulating a plasma disruption in a thermonuclear fusion reactor. The test pieces were heated by an intense hydrogen ion beam. The heated area was about 70 mm in diameter. The peak heat flux on the surface ranged from 68 to 261 MW/m 2 , and the heating duration from 40 to 250 ms. The melting and evaporating process was observed by using a high-speed video camera. The melt layer convected from the center of the piece to the periphery and the thickness of the piece was decreased not only by the evaporation but also by the convection in the melt layer.

High heat flux experiments at JAERI

Recent R&D results on high-heat flux components are presented, including construction of a new test stand. The test stand can extract an electron beam of 4.1 at 100 keV. E-folding divergence of the beam is 1.7 mrad, and the latest beam performance is also described. At the original test stand, which can produce hydrogen-ion beams of 50 A at 100 keV for 10 s, high-Z divertor armors, were tested. Tungsten plates brazed on copper blocks have been proven to have enough durability against heat flux under 10 MW/m/sup 2/. Carbon-fiber-carbon (CFC) composites were tested at the new electron-beam test stand and an electron-beam welding machine. Under disruption-simulation conditions, evaporation weight loss of CFC was lower than that of isotropic graphite.<<ETX>>

Concept of an inherently-safe high temperature gas-cooled reactor

As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of...

Research and development on system integration technology for connection of hydrogen production system to an HTGR

The Japan Atomic Energy Agency (JAEA) has been promoting research and development on the hydrogen production technology with a high-temperature gas-cooled reactor (HTGR), with a view to contributing to the global warming issue and hydrogen energy society in the near future. The system integration technology for connection of the hydrogen production system to the HTGR is one of the key technologies to put hydrogen production with nuclear energy to commercial use. Research and development on the system integration technology has been carried out for four items: control technology to maintain reactor operation against thermal disturbance caused by the hydrogen production system, estimation of the tritium permeation into the hydrogen from the reactor, a countermeasure against explosion, and development of a high-temperature valve to isolate the reactor and hydrogen production systems in accidents. This report describes the research activities on the system integration technology at JAEA.

Performance Test Results of Mock-up Test Facility of HTTR Hydrogen Production System

For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m3/h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004.

A thermal cycling test of tungsten copper bonds for divertor collector plates

A tungsten-copper (W-Cu) bond is a reference design concept for a divertor collector plate in the Fusion Experimental Reactor (FER). In the present study, durability of the W-Cu bond under thermal cycling was experimentally examined. A plasma spraying technique was applied to form a tungsten layer with a thickness of around 2 mm on a copper disk. This technique has advantages in that tungsten can be deposited on curved and complex surfaces of substrates and that the temperature of the substrate can be kept low during the deposition. Two kinds of copper materials, oxygen-free high conductivity copper (Cu/OFHC) and copper alloyed with beryllium (Cu/Be), were used. The tungsten surface of the W-Cu test piece was periodically irradiated with a high temperature argon plasma jet under the thermal stress condition corresponding to that expected at the normal operation of the FER. The thermal properties of the plasma sprayed tungsten were measured before the test, and the microstructure before and after the test was investigated with a scanning electron microscope. The sprayed tungsten layer was found stratified and had porosity of 2–3 % before the test. The W-Cu/OFHC test piece survived for thermal cycle up to 6000, while the W-Cu/Be test piece broke between 4500 and 5000 cycles.

A thermal cycling durability test of tungsten copper duplex structures for use as a divertor plate

A tungsten-copper (W—Cu) duplex structure is the reference divertor plate design concept for the Fusion Experimental Reactor (FER). In the present study, a durability test of the W—Cu duplex structure against thermal cycling was performed. A tungsten disk was bonded to a copper disk by means of brazing or direct casting. The tungsten surface of the test piece was periodically heated by a high temperature argon plasma jet. Before and after the tests, the test pieces were examined with the aid of a scanning electron microscope and the Knoop hardness was measured. Grain boundary microcracks were observed after 200 and 1100 thermal cycles in brazed tungsten samples which contained a small amount of nickel and phosphorus. A cast tungsten specimen subjected to 2200 thermal cycles also contained microcracks. However, microcracks were not observed in a brazed tungsten sample containing an extremely small amount of impurities for thermal cycles up to 3700 times. Microcracks were observed in the brazing material of this test piece. None of the test specimens were broken. It is found that brazing is a valid bonding method and that W—Cu duplex structure, especially with high purity tungsten, is able to endure a practical number of thermal cycles.