Feasibility study on the blanket tritium recovery system using the palladium membrane diffuser

Kawamura, Yoshinori; Enoeda, Mikio; Yamanishi, Toshihiko; Nishi, Masataka

Fusion Engineering and Design, 81(1-7), p.809 - 814, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.09.035

Tritium bred in the solid breeder blanket of a fusion reactor is extracted by passing of a helium sweep gas. Tritium is separated from sweep gas at the blanket tritium recovery system. Palladium membrane diffuser is one of the applicable processes for the blanket tritium recovery system. It is usually applied for hydrogen purification system such as TEP in ITER. However, it has been reported that the rate controlling step changes at lower hydrogen pressure such as the blanket sweep gas condition, and discussion about application for the blanket sweep gas condition is not enough. Recently, conceptual design of the demonstration reactor, named "DEMO2001", has been proposed from JAERI. In this report, the application of the Pd diffuser for the blanket sweep gas condition is discussed based on the condition of DEMO 2001.

Evaluation of tritium behavior in the epoxy painted concrete wall of ITER hot cell

Nakamura, Hirofumi; Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

Fusion Science and Technology, 48(1), p.452 - 455, 2005/07

https://doi.org/10.13182/FST05-A963

Tritium behavior released in ITER hot cell has been investigated numerically. Tritium behavior was evaluated by a combined analytical methods of a tritium transport analysis with the one dimensional diffusion model in the multi-layer wall (concrete and epoxy paint) and a tritium concentration analysis with the complete mixing model by the ventilation in the hot cell under the simulated hot cell operational conditions. As the results, tritium concentration in the hot cell volume decreases rapidly from 300 DAC (Derived Air Concentration) less than 1 DAC in several days after removing the tritium release source. Tritium inventory in the wall is estimated to be about 0.1 PBq for 20 years operation. On the other hand, Tritium permeation through the epoxy painted concrete wall will be negligible. Finally, as to the effect of epoxy paint on the tritium permeation and inventory, it is found that the epoxy paint can reduce tritium inventory by about two orders of magnitude relative to bare concrete wall.

Evaluation of permeated hydrogen through heat transfer pipes of the intermediate heat exchanger during the initial 950C operation of the HTTR

Sakaba, Nariaki; Matsuzawa, Takaharu*; Hirayama, Yoshiaki*; Nakagawa, Shigeaki; Nishihara, Tetsuo; Takeda, Tetsuaki

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 8 Pages, 2005/05

The permeation of hydrogen isotopes through the Hastelloy XR high-temperature alloy adopted for the heat exchanger pipes of the intermediate heat exchanger in the HTTR (High Temperature Engineering Test Reactor) is one of the concerns in the hydrogen production system, which will be connected to the HTTR in the near future. An evaluation of the hydrogen permeation between the primary and secondary coolant through the Hastelloy XR was performed using the hydrogen concentration data observed during the initial 950C operation of the HTTR. The hydrogen permeability of the Hastelloy XR was estimated conservatively high as follows. The activation energy E and pre-exponential factor F of the permeability of hydrogen were E = 62 kJ/mol and F = 3.610 cm(NTP)/(cm s Pa), respectively, in the temperature range from 735K to 940K. The results implied that some oxidized film had been formed on the surface of the heat exchanger pipes of the intermediate heat exchanger.

Hydrogen permeation measurement of the reduced activation ferritic steel F82H by the vacuum thermo-balance method

Yoshida, Hajime; Kosaku, Yasuo*; Enoeda, Mikio; Abe, Tetsuya; Akiba, Masato

JAERI-Research 2005-003, 13 Pages, 2005/03

JAERI-Research-2005-003.pdf:3.33MB

Hydrogen permeation fluxes of the reduced activation ferritic steel F82H were quantitatively measured by a newly proposed method, vacuum thermo-balance method, for a precise estimation of tritium leakage in a fusion reactor. We prepared sample capsules made of F82H, which enclosed hydrogen gas. The hydrogen in the capsules permeated through the capsule wall, and subsequently desorbed from the capsule surface during isothermal heating. The vacuum thermo-balance method allows simultaneous measurement of the hydrogen permeation flux by two independent methods, namely, the net weight reduction of the sample capsule and exhaust gas analysis. Thus the simultaneous measurements by two independent methods increase the reliability of the permeability measurement. The ratio of the hydrogen permeation fluxes obtained by the net weight reduction to that measured by the exhaust gas analysis was in the range from 1/4 to 1/1 in this experiment. It has been demonstrated that the vacuum thermo-balance method is effective for the measurement of hydrogen permeation rate of F82H.

Counter-permeation of deuterium and hydrogen through INCONEL 600

Takeda, Tetsuaki; Iwatsuki, Jin*

Nuclear Technology, 146(1), p.83 - 95, 2004/04

https://doi.org/10.13182/NT04-A3490

The objective of this study is to investigate the effect of the existence of hydrogen in a pipe outside on the amount of permeated deuterium through the pipe. It was found that the amount of permeated deuterium decreases with increasing the partial pressure of hydrogen in the pipe outside when the partial pressure of deuterium in the pipe is lower than 100 Pa and that of hydrogen in the pipe outside is higher than 10 kPa. The amount of permeated deuterium on counter permeation was predicted quantitatively by using an effectiveness factor for diffusivity of deuterium in metals and by taking into account the equilibrium state for hydrogen, deuterium and HD molecules on the metal surface. From the results obtained in this study, it is supposed that the amount of tritium transferred from the primary circuit of the HTTR to the hydrogen production system will be reduced by the existence of high-pressure hydrogen in the catalyst pipe of the steam reformer.

Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

Nakamura, Hirofumi; Nishi, Masataka

JAERI-Research 2003-024, 24 Pages, 2003/11

JAERI-Research-2003-024.pdf:1.12MB

Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium transport properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authors' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evaluation was also carried out for comparison (previous data). The permeation analysis was carried out individually by classifying into the armor region (Carbon Fiber Composites and tungsten) and the slit region without armor (3% of armor surface area) assuming the incident flux and temperature for each region. As the results of the permeation analysis, estimated permeation amount with the authors' data was one order less than that with the previous data at the end of lifetime of the divertor due to authors' small diffusion coefficient of tritium in tungsten. It also indicated the possibility that permeation through the slit region of the armor tiles could dominate total permeation through the vertical target, since tritium permeation amount through tungsten armor with the authors' data was estimated to be reduced drastically smaller than that with the previous evaluation data. The result of a little tritium permeation amount through the vertical target with the authors' data ensured the conservatism of the current evaluation of tritium concentration in the primary cooling water in ITER divertor, as it indicated the possibility of direct drainage of the divertor primary cooling water.

Conceptual design of solid breeder blanket system cooled by supercritical water

Enoeda, Mikio; Ohara, Yoshihiro; Akiba, Masato; Sato, Satoshi; Hatano, Toshihisa; Kosaku, Yasuo; Kuroda, Toshimasa*; Kikuchi, Shigeto*; Yanagi, Yoshihiko*; Konishi, Satoshi; et al.

JAERI-Tech 2001-078, 120 Pages, 2001/12

JAERI-Tech-2001-078.pdf:8.3MB

This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. This conceptual design study was performed to determine the updated strategy and goal of the R&D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology.

Implantation driven permeation behavior of deuterium through pure tungsten

Nakamura, Hirofumi; Hayashi, Takumi; Nishi, Masataka; Arita, Makoto; Okuno, Kenji*

Fusion Engineering and Design, 55(4), p.513 - 520, 2001/09

https://doi.org/10.1016/S0920-3796(01)00380-5

no abstracts in English

Evaluation tritium transportation to the product hydrogen in the HTGR hydrogen production system

Nishihara, Tetsuo; Hada, Kazuhiko

Nihon Genshiryoku Gakkai-Shi, 41(5), p.571 - 578, 1999/05

https://doi.org/10.3327/jaesj.41.571

no abstracts in English

Measurement of HTO permeability of materials for protective appliances

Yamamoto, Hideaki; *; Kato, Shohei; Murata, Mikio; Kinouchi, Nobuyuki;

Proc. of the Int. Radiation Protection Association,Vol. 1, p.467 - 470, 1992/00

no abstracts in English