Satety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder

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概要

• 論文の詳細を見る

• Performance test of test blanket modules in the fusion environment using the International Thermonuclear Experimental Reactor (ITER) is one of the most important mile-stone for the development of the breeding blanket of the fusion power plant. In the design of test blanket modules in the ITER, it is very important to show that test modules do not cause additional safety concern to the ITER. This work has been performed for the evaluation of the preliminary safety of the test blanket module of a water cooled solid blanket, which is the primary candidate of the breeding blanket in Japan currently. Major issues of the evaluation were, establishment of post-accident cooling in the test blanket module, hydrogen gas generation by Be/steam reaction, and pressure increase and spilled water amount by the event of coolant leakage. The analyses results showed that, suppression tank system is necessary to accommodate the over-pressure by the coolant water after pipe break in the box of the test module. Coolant water pipe break of the first wall of the test blanket module will result relatively small impact to the ITER safety because of the small inventory of the coolant water of the test module and large volume of the vacuum vessel of the ITER. However, it was clarified that the water cooled blanket with beryllium pebble as the multiplier will have the potential hazard of the hydrogen formation. Further investigation to maintain the safety on this aspect is required.

• 社団法人 日本原子力学会の論文
• 2001-11-25

著者

• Moriyama Koichi

  Kawasaki Heavy Industries Ltd.

• Ohara Yoshihiro

  Kawasaki Heavy Industries Ltd.

• Enoeda M

  Japan Atomic Energy Res. Inst. Ibaraki Jpn

• ENOEDA Mikio

  Blanket Engineering Laboratory, Japan Atomic Energy Research Institute

• KURODA Toshimasa

  Blanket Engineering Laboratory, Japan Atomic Energy Research Institute

• Kuroda Toshimasa

  Blanket Engineering Laboratory Japan Atomic Energy Research Institute

関連論文

• Low Cycle Fatigue Lifetime of HIP Bonded Bi-metallic First Wall Structures of Fusion Reactors
• Nuclear and Thermal Analyses of Supercritical-water-cooled Solid Breeder Blanket for Fusion DEMO Reactor
• Satety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder
• MULTILAYER ADSORPTION CHARACTERISTICS OF HEAVY WATER ON ACTIVATED ALUMINA
• A MULTILAYER MODEL FOR ADSORPTION OF WATER ON ACTIVATED ALUMINA IN RELATION TO ADSORPTION POTENTIAL

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