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Journal Articles

Nuclear and thermal feasibility of lithium-loaded high temperature gas-cooled reactor for tritium production for fusion reactors

Goto, Minoru; Okumura, Keisuke; Nakagawa, Shigeaki; Inaba, Yoshitomo; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*

Fusion Engineering and Design, 136(Part A), p.357 - 361, 2018/11

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

A High Temperature Gas-cooled Reactor (HTGR) is proposed as a tritium production device, which has the potential to produce a large amount of tritium using $$^{6}$$Li(n,$$alpha$$)T reaction. In the HTGR design, generally, boron is loaded into the core as a burnable poison to suppress excess reactivity. In this study, lithium is loaded into the HTGR core instead of boron and is used as a burnable poison aiming to produce thermal energy and tritium simultaneously. The nuclear characteristics and the fuel temperature were calculated to confirm the feasibility of the lithium-loaded HTGR. It was shown that the calculation results satisfied the design requirements and hence the feasibility was confirmed for the lithium-loaded HTGR, which produce thermal energy and tritium.

Journal Articles

Evaluation of tritium confinement performance of alumina and zirconium for tritium production in a high-temperature gas-cooled reactor for fusion reactors

Katayama, Kazunari*; Ushida, Hiroki*; Matsuura, Hideaki*; Fukada, Satoshi*; Goto, Minoru; Nakagawa, Shigeaki

Fusion Science and Technology, 68(3), p.662 - 668, 2015/10

 Times Cited Count:7 Percentile:31.87(Nuclear Science & Technology)

Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety and production efficiency, tritium confinement technique is an important issue. It is known that alumina has high resistance for gas permeation. In this study, hydrogen permeation experiments in commercial alumina tubes were conducted and hydrogen permeability, diffusivity and solubility was evaluated. By using obtained data, tritium permeation behavior from an Al$$_{2}$$O$$_{3}$$-coated Li-compound particle was simulated. Additionally, by using literature data for hydrogen behavior in zirconium, an effect of Zr incorporation into an Al$$_{2}$$O$$_{3}$$ coating on tritium permeation was discussed. It was indicated that the majority of produced tritium was released through the Al$$_{2}$$O$$_{3}$$ coating above 500$$^{circ}$$C. However, it is expected that total tritium leak is suppressed to below 0.67% of total tritium produced at 500$$^{circ}$$C by incorporating Zr fine particles into the inside of Al$$_{2}$$O$$_{3}$$ coating.

Journal Articles

Study on operation scenario of tritium production for a fusion reactor using a high temperature gas-cooled reactor

Kawamoto, Yasuko*; Nakaya, Hiroyuki*; Matsuura, Hideaki*; Katayama, Kazunari*; Goto, Minoru; Nakagawa, Shigeaki

Fusion Science and Technology, 68(2), p.397 - 401, 2015/09

 Times Cited Count:1 Percentile:85.75(Nuclear Science & Technology)

To start up a fusion reactor, it is necessary to provide a sufficient amount of tritium from an external device. Herein, methods for supplying a fusion reactor with tritium are discussed. Use of a high temperature gas cooled reactor (HTGR) as a tritium production device has been proposed. So far, the analyses have been focused only on the operation in which fuel is periodically exchanged (batch) using the block type HTGR. In the pebble bed type HTGR, it is possible to design an operation that has no time loss for refueling. The pebble bed type HTGR (PBMR) and the block type HTGR (GTHTR300) are assumed as the calculation and comparison targets. Simulation is made using the continuous-energy Monte Carlo transport code MVPBURN. It is shown that the continuous operation using the pebble bed type HTGR has almost the same tritium productivity compared with the batch operation using the block type HGTR. The issues for pebble bed type HTGR as a tritium production device are discussed.

Journal Articles

Application of the high-temperature gas-cooled reactor to produce tritium for fusion reactors

Nakaya, Hiroyuki*; Matsuura, Hideaki*; Katayama, Kazunari*; Goto, Minoru; Nakagawa, Shigeaki

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.398 - 402, 2015/05

The performance of tritium production for fusion reactor using High-Temperature Gas-cooled Reactor (HTGR) is studied. An influence of $$^{6}$$Li concentration on tritium production performance using HTGR is estimated. Li compound is loaded in the reactor core using Li rod consisting cylindrical Li compound in cladding tube. A Gas Turbine High-Temperature Reactor of 300 MWe nominal capacity (GTHTR300) with 600 MW thermal output power is assumed as HTGR. An amount of tritium production is estimated by burn-up calculations using the continuous-energy Monte Carlo transport code MVP-BURN. The amount of tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. Even if 6Li is enriched, the GTHTR300 can produce 500 g of tritium over 180-day operation without increasing the amount of required $$^{6}$$Li. The amount of tritium outflow is decreased by 20-50%.

Journal Articles

Correlation of rates of tritium migration through porous concrete

Fukada, Satoshi*; Katayama, Kazunari*; Takeishi, Toshiharu*; Edao, Yuki; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 67(2), p.99 - 102, 2015/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Overview of plasma-material interaction experiments on EAST employing MAPES

Ding, F.*; Luo, G.-N.*; Pitts, R.*; Litnovsky, A.*; Gong, X.*; Ding, R.*; Mao, H.*; Zhou, H.*; Wampler, W. R.*; Stangeby, P. C.*; et al.

Journal of Nuclear Materials, 455(1-3), p.710 - 716, 2014/12

 Times Cited Count:20 Percentile:8.06(Materials Science, Multidisciplinary)

Journal Articles

Transfer of tritium in concrete coated with hydrophobic paints

Fukada, Satoshi*; Edao, Yuki*; Sato, Koichi*; Takeishi, Toshiharu*; Katayama, Kazunari*; Kobayashi, Kazuhiro; Hayashi, Takumi; Yamanishi, Toshihiko; Hatano, Yuji*; Taguchi, Akira*; et al.

Fusion Engineering and Design, 87(1), p.54 - 60, 2012/01

 Times Cited Count:4 Percentile:60.31(Nuclear Science & Technology)

An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (1) how fast HTO penetrates through concrete walls, (2) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (3) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The epoxy paint coating can work as a HTO diffusion barrier and the PRF value is around 1/10. The silicon paint coating cannot work as the anti-T permeation barrier, because water deteriorates contact between the paint and cement or mortar.

Journal Articles

Concentration profiles of tritium penetrated into concrete

Takata, Hiroki*; Furuichi, Kazuya*; Nishikawa, Masabumi*; Fukada, Satoshi*; Katayama, Kazunari*; Takeishi, Toshiharu*; Kobayashi, Kazuhiro; Hayashi, Takumi; Namba, Haruyuki*

Fusion Science and Technology, 54(1), p.223 - 226, 2008/07

 Times Cited Count:8 Percentile:45.69(Nuclear Science & Technology)

Concentration profiles of tritium penetrated into cement paste, mortar and concrete were measured by using samples with a shape of column. Tritium penetrated until a location of about 5 cm from the exposed surface after 6 months' exposure. The amount of tritium penetrated into mortar and concrete were less than 70% and half that into cement paste.

Journal Articles

Tritium release from bulk of carbon-based tiles used in JT-60U

Takeishi, Toshiharu*; Katayama, Kazunari*; Nishikawa, Masabumi*; Masaki, Kei; Miya, Naoyuki

Journal of Nuclear Materials, 349(3), p.327 - 338, 2006/03

 Times Cited Count:6 Percentile:55.07(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Release behavior of hydrogen isotopes from JT-60U graphite tiles

Katayama, Kazunari*; Takeishi, Toshiharu*; Nagase, Hiroyasu*; Manabe, Yusuke*; Nishikawa, Masabumi*; Miya, Naoyuki; Masaki, Kei

Fusion Science and Technology, 48(1), p.561 - 564, 2005/07

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Recovery of retained tritium from graphite tile of JT-60U

Takeishi, Toshiharu*; Katayama, Kazunari*; Nishikawa, Masabumi*; Miya, Naoyuki; Masaki, Kei

Fusion Science and Technology, 48(1), p.565 - 568, 2005/07

 Times Cited Count:1 Percentile:88.49(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Katayama, Kazunari*; Takeishi, Toshiharu*; Manabe, Yusuke*; Nagase, Hiroyasu*; Nishikawa, Masabumi*; Miya, Naoyuki

Journal of Nuclear Materials, 340(1), p.83 - 92, 2005/04

 Times Cited Count:6 Percentile:56.05(Materials Science, Multidisciplinary)

no abstracts in English

Oral presentation

Engineering study of tritium production with high temperature gas-cooled reactor

Goto, Minoru; Nakagawa, Shigeaki; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*

no journal, , 

A high temperature gas-cooled reactor (HTGR) is proposed as a tritium production device, which can produce a large amount of tritium by loading coated Li pebbles into the core without a major change of the original reactor core design. The engineering study was performed to clarify the problem of the tritium production system and assess the possibility of the resolution. As a result, it was assessed that all of the problems could be resolved from an engineering view point by using the HTTR (High Temperature engineering Test Reactor) experiences on the production of the coated fuel particles and deal of the fuels.

Oral presentation

Study on tritium production with high-temperature gas-cooled reactor for fusion reactors, 2; Process and plant safety

Goto, Minoru; Nakagawa, Shigeaki; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*

no journal, , 

The tritium fuel production system using a high-temperature gas-cooled reactor is proposed for initial fusion reactors. In this study, the feasibility assessment of the system was performed from the view point of engineering and safety.

Oral presentation

The Influence of gap geometry on impurity deposition and fuel accumulation in the castellated tungsten plasma-facing components exposed in EAST

Ding, F.*; Ashikawa, Naoko*; Fukumoto, Masakatsu; Katayama, Kazunari*; Mao, H.*; Ding, R.*; Xu, Q.*; Wu, J.*; Xie, C. Y.*; Luo, G.-N.*

no journal, , 

Oral presentation

Evaluation of nuclear characteristics of lithium loaded HTGR for tritium production for fusion reactors

Goto, Minoru; Okumura, Keisuke; Nakagawa, Shigeaki; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*

no journal, , 

A feasibility study of a High Temperature Gas-cooled Reactor (HTGR) for tritium production using $$^{6}$$Li(n,$$alpha$$)T reaction for fusion reactors has been conducted. In this study, the burn-up chain was modified to treat $$^{6}$$Li(n,a)T reaction directory in neutronics calculations, and then the feasibility study was performed from the view point of nuclear characteristics using SRAC code system, which has experience in neutronics analysis of HTGRs.

Oral presentation

Study on tritium production for initial fusion reactor using high temperature gas cooled reactor; Improvement of the tritium containment performance by concentrating Li compound

Nakaya, Hiroyuki*; Matsuura, Hideaki*; Kawamoto, Yasuko*; Nagasumi, Satoru*; Katayama, Kazunari*; Goto, Minoru; Nakagawa, Shigeaki

no journal, , 

We proposed the used of High Temperature Gas-cooled Reactors (HTGR) as a tritium production device, which produces tritium by $$^{6}$$Li(n,$$alpha$$)T reaction, for initial fusion reactors. Concentrating of $$^{6}$$Li suppresses undesirable leakage of produced tritium into reactor coolant. In this study, the effect of $$^{6}$$Li concentration difference on the amount of the tritium leakage and the tritium production efficiency was investigated.

Oral presentation

Study on management of tritiated water for a fusion DEMO reactor

Watanabe, Kazuhito; Nakamura, Makoto; Someya, Yoji; Masui, Akihiro; Katayama, Kazunari*; Hayashi, Takumi; Yanagihara, Satoshi*; Konishi, Satoshi*; Yokomine, Takehiko*; Torikai, Yuji*; et al.

no journal, , 

In the DEMO design, the blanket primary cooling system involves high temperature pressurized water (~300$$^{circ}$$C). This means the temperature of blanket structural material is higher than that of ITER. This increases tritium permeation ratio from the fusion plasma and blanket breeder to the primary cooling water. Therefore, we need to consider installation of a water detritiation system. In this study, we estimate the demand of water detritiation system from the view point of the amount of tritium permeated to primary cooling water that assumed conservatively. We also organize the issues for management of tritiated water from the other point of view based on the characteristic of the fusion DEMO reactor. The result shows that the existing facilities can be adopted to the DEMO if we can control the tritium ratio of primary cooling water as same as that of CANDU reactor.

Oral presentation

Evaluation of nuclear and thermal characteristics of lithium loaded HTGR for tritium production

Goto, Minoru; Okumura, Keisuke; Nakagawa, Shigeaki; Inaba, Yoshitomo; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*

no journal, , 

A feasibility study of a High Temperature Gas-cooled Reactor (HTGR) for tritium production using $$^{6}$$Li(n,$$alpha$$)T reaction for fusion reactors has been conducted. In this study, the burn-up chain was modified to treat $$^{6}$$Li(n,$$alpha$$)T reaction directory in neutronics calculations, and then the feasibility study was performed from the view point of nuclear and thermal characteristics using SRAC code system, which has experience in neutronics analysis of HTGRs.

Oral presentation

Tritium production and containment performance of lithium rod test module for high temperature gas-cooled reactor

Ida, Yuma*; Matsuura, Hideaki*; Nagasumi, Satoru*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki

no journal, , 

JAEA and Kyushu University have studied the tritium production method using high temperature gas-cooled reactors (HTGR) for initial fusion reactors. In this method, lithium compounds are loaded into the reactor core and tritium is produced with $$^{6}$$Li(n,$$alpha$$)T reaction. We studied about optimization of lithium loading method, effective tritium containment method and nuclear thermal design of lithium loaded HTGR, and consequently we confirmed the feasibility of the tritium production method. Then, we started preliminary study for lithium irradiation experiment by test reactors. This paper describes evaluation results of tritium production and tritium containment for proposed lithium irradiation capsule.

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