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Tanigawa, Hisashi; Gwon, H.; Kawamura, Yoshinori
no journal, ,
JAEA is developing blanket with a water-cooled ceramic breeder concept. The blanket has a box structure made of reduced-activation ferritic/martensitic steels, and pebbles of tritium breeding and neutron multiplier materials are packed into the box. To remove nuclear heating and surface heat flux, the box structure has built-in cooling paths and cooling pipes are inserted into the packed pebble beds. The configuration of these materials has to be designed so that both necessity tritium capability and allowable material temperatures are kept. A two-dimensional nuclear-thermal-coupled analysis code has been developed for the design activity. The code gives us decay heat distribution and its time dependency in the blanket in addition to the nuclear and thermal property of the blanket under normal operation. Functions of the code and procedure how to use the obtained values in the design are reported.
Gwon, H.; Tanigawa, Hisashi; Hirose, Takanori; Kawamura, Yoshinori
no journal, ,
Thermal structural characteristics for Test Blanket Module (TBM) of a Water Cooled Ceramic Breeder (WCCB) were evaluated by using a Finite Element Method (FEM). A single load conditions such as surface heat flux, nuclear heating, and pressure of coolant were applied as boundary conditions. Thermal stress that was caused by the heat loads and the internal pressure was evaluated. In addition the combined load conditions of the single load conditions were also considered. Based on criteria corresponding to the different loads, the integrity of TBM was evaluated.
Kawamura, Yoshinori; Hirose, Takanori; Tanigawa, Hisashi; Nakajima, Motoki; Gwon, H.; Miyata, Satoru; Sato, Satoshi
no journal, ,
Japan Domestic Agency (JADA) participates to the Test Blanket Module (TBM) test program of the ITER project as the lead party of Water Cooled Ceramic Breeder (WCCB) - TBM. We have signed TBM arrangement (TBMA) in Nov. 2014, and conceptual design has been reviewed in Feb. 2015. We have received 3 Cat-1 chits. Solution of Cat-1 chit is necessary to advance to the next design phase (preliminary design). Other intrinsic topic is the study on flow-accelerated corrosion of F82H. F82H is one of Reduced Activation Ferritic/Martensitic steels (RAFM) developed in Japan. In the piping of water cooling system, both F82H and stainless steel are used. As for fission reactor, there are many reports about corrosion of stainless steel. However, the report about F82H, especially flow-accelerated corrosion, is few. So, in this report, we will mainly show the situation for chit resolution, and the progress of the study on flow-accelerated corrosion of F82H.
Gwon, H.; Tanigawa, Hisashi; Nakajima, Motoki; Hirose, Takanori; Kawamura, Yoshinori
no journal, ,
It is concerned that the temperature of the blanket would increase excessively due to the decay heat even after the plasma shutdown. In present study, we focuses the relationship between the blanket structures and the passive cooling performance and considers how to effectively mitigate the excessive temperature rising due to the decay heat. The arrangement of both pebble beds and the ribs was changed as a way to relieve the temperature rising. The thermal response characteristics of the modified blanket models were evaluated by using a two-dimensional nuclear-thermal-coupled analysis code, DOHEAT3. The TBR of each model was also evaluated. Based on the results, the useful design policies for blanket were proposed from the viewpoint of the decay heat removal.
Nakajima, Motoki; Hirose, Takanori; Gwon, H.; Tanigawa, Hisashi; Kawamura, Yoshinori
no journal, ,
As the primary candidate of ITER-TBM of Japan, development of WCCB TBM is being performed. In this blanket, the high temperature pressurized water was used as the coolant. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. Additionally, it is also required that understanding of corrosion properties of temperature ranging from 543K to 593K, because inlet and outlet temperatures of TBM are 543K and 593K, respectively. In this study, the corrosion test was performed at temperature ranging from 543K to 593K in deaerated (20ppb DO) and oxygen-saturated (8ppm DO) high temperature water.
Tanigawa, Hisashi; Gwon, H.; Hirose, Takanori; Nakajima, Motoki; Kawamura, Yoshinori
no journal, ,
Blanket in fusion nuclear reactor has three functions such as neutron shielding, heat recovery and tritium breeding. Blankets with ceramic breeder materials has a box structure made of reduced activation ferritic/martensitic steel, and pebbles of breeder and neutron multiplier materials are packed into the box. Heat and neutron loads on the blanket are cooled by high temperature and pressure coolant such as water and helium. In this study, strength and pressure integrity are assessed for the water-cooled ceramic breeder blanket developed in JAEA. Existing design standards for pressure equipment are used in the analysis and their applicability to the fusion blanket is discussed.
Hirose, Takanori; Tanigawa, Hisashi; Nakajima, Motoki; Gwon, H.; Miyata, Satoru; Sato, Satoshi; Kawamura, Yoshinori; Yamanishi, Toshihiko
no journal, ,
no abstracts in English
Gwon, H.; Tanigawa, Hisashi; Nakajima, Motoki; Hirose, Takanori; Kawamura, Yoshinori
no journal, ,
It is expected that the neutron wall loading in DEMO is larger than that in ITER, over 0.78 MW/m
. It is concerned that the decay heat due to the large neutron wall loading will lead to excessive temperature rising in blanket. In present study, the thermal response characteristics of blanket caused by the decay heat under LOCA were evaluated. In addition we considered how to effectively mitigate the excessive temperature rising based on the evaluation results.
Tanigawa, Hisashi; Gwon, H.; Kawamura, Yoshinori
no journal, ,
JAEA is developing a water-cooled ceramic breeder blanket. For the blanket strength and pressure integrity are assessed. The largest stress appears in the first wall region due to the surface heat and neutron loads. Under conditions with the thermal loads and cooling water pressure, stress in the first wall is analyzed with reference to ASME Boiler and Pressure Vessel Code. Limitations related to the primary/secondary stresses and strain are considered. Material data necessary for the assessment is summarized, and then status of preparation is studied for a structural material of reduced activation ferritic/martensitic steel, F82H. The summarized date of F82H is compared with standardized 9Cr-1Mo-V.
Gwon, H.; Tanigawa, Hisashi; Hirose, Takanori; Kawamura, Yoshinori
no journal, ,
Design/Manufacturing of Test Blanket Module (TBM), TBM test for demonstration of major functions of blanket are one of the most important subjects which are closely related with Design/Manufacturing of blanket in prototype reactor. In addition to that it is necessary to investigate the design which can cover load condition of prototype reactor. Unfortunately, however, the examination concerning the load condition of prototype reactor was still insufficient. In present study, we focused on pressure rise in blanket box as well as increase of decay heat under In Box LOCA. The characteristics of thermal structural response of blanket was investigated. Based on the results we proposed the design which can handle the conditions of prototype reactor.