@article{oai:repo.qst.go.jp:00075612,
 author = {Hiwatari, Ryoji and Katayama, Kazunari and Nakamura, Makoto and Miyoshi, Yuuya and Aoki, Akira and Asakura, Nobuyuki and Uto, Hiroyasu and Homma, Yuuki and Tokunaga, Shinsuke and Nakajima, Noriyasu and Someya, Yoji and Sakamoto, Yoshiteru and Tobita, Kenji and Special Design Team for Fusion DEMO, Joint and Hiwatari, Ryoji and Nakamura, Makoto and Miyoshi, Yuuya and Asakura, Nobuyuki and Uto, Hiroyasu and Homma, Yuuki and Tokunaga, Shinsuke and Someya, Yoji and Sakamoto, Yoshiteru and Tobita, Kenji},
 issue = {C},
 journal = {Fusion Engineering and Design},
 month = {Apr},
 note = {The conceptual design of Japan’s fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in- vessel components to the primary cooling system is evaluated to be 5.7g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 318 Ci/year/loop, which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, confinement concept of tritium release from PHTS is discussed under the condition of an ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA.},
 pages = {259--266},
 title = {Development of plant concept related to tritium handling in the water-cooling system for JA DEMO},
 volume = {143},
 year = {2019}
}