Science from the initial operation of HRC

Ito, Shinichi*; Yokoo, Tetsuya*; Masuda, Takatsugu*; Yoshizawa, Hideki*; Soda, Minoru*; Ikeda, Yoichi*; Ibuka, Soshi*; Kawana, Daichi*; Sato, Taku*; Nambu, Yusuke*; et al.

JPS Conference Proceedings (Internet), 8, p.034001_1 - 034001_6, 2015/09

https://doi.org/10.7566/JPSCP.8.034001

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki

Nuclear Engineering and Design, 271, p.309 - 313, 2014/05

https://doi.org/10.1016/j.nucengdes.2013.11.052

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. The air/water ingress accident, one of the DECs for the HTGR, is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

Evaluation of fracture toughness of fine-grained isotropic graphites for HTGR

Yamada, Teruaki*; Matsushima, Yuki*; Kuroda, Masatoshi*; Sumita, Junya; Shibata, Taiju; Fujita, Ichiro*; Sawa, Kazuhiro

Nuclear Engineering and Design, 271, p.323 - 326, 2014/05

https://doi.org/10.1016/j.nucengdes.2013.11.055

In order to investigate the effects of the experimental methodology and the notch angle on the fracture toughness of the fine-grained isotropic nuclear graphites IG-110 and IG-430, the three-point-bending test, which has been recently proposed as the methodology to evaluate the fracture toughness of graphite for high temperature gas-cooled reactors (HTGRs), was performed using two types of the specimens with different notch angles. The results obtained in this study could be summarized as follows: (1) The values of the fracture toughness of IG-110 and IG-430 measured in this study were 0.890 MPa m and 1.031 MPa m, respectively. It was also found that the value of the fracture toughness of IG-110 was nearly equal to or smaller than the values obtained by the other method reported previously. (2) The values of the fracture toughness of the fine-grained isotropic graphites were not affected between the notch angles introduced by the incisive razor blade. (3) The ratio of the tensile strengths of IG-110 and IG-430 was estimated from Griffith Theory using the experimental data obtained in this study. The estimated strength ratio was in good agreement with the strength ratio obtained from the supplier's data.

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2012

Sumiya, Shuichi; Watanabe, Hitoshi; Miyagawa, Naoto; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Inoue, Kazumi; Yoshii, Hideki; Otani, Kazunori*; Hiyama, Yoshinori*; et al.

JAEA-Review 2013-041, 115 Pages, 2014/01

JAEA-Review-2013-041.pdf:19.01MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, and the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki prefecture), this report describes the effluent control results of liquid waste discharged from the JAEA's Nuclear Fuel Cycle Engineering Laboratories in the fiscal year 2012, from 1st April 2012 to 31st March 2013. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other facilities were much lower than the authorized limits of the above regulations.

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2011

Sumiya, Shuichi; Watanabe, Hitoshi; Miyagawa, Naoto; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Hiyama, Yoshinori; Yoshii, Hideki*; Otani, Kazunori*; Goto, Ichiro*; et al.

JAEA-Review 2013-005, 116 Pages, 2013/05

JAEA-Review-2013-005.pdf:9.69MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, and the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and byelaw of Ibaraki prefecture), this report describes the effluent control results of liquid waste discharged from the JAEA's Nuclear Fuel Cycle Engineering Laboratories in the fiscal year 2011, from 1st April 2011 to 31st March 2012. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other facilities were much lower than the authorized limits of the above regulations.

Development of grout materials for a geological disposal system for high-level radioactive waste, 2

Kawaguchi, Masanao; Nakanishi, Tatsuro; Kishi, Hirokazu; Nobuto, Jun*; Yamada, Tsutomu*; Fujita, Tomoo; Hatanaka, Koichiro

JAEA-Data/Code 2012-007, 250 Pages, 2012/11

JAEA-Data-Code-2012-007.pdf:20.55MB

Cementitious materials are commonly used for rock support, lining, and grouting, their pH plume are considered to have an adverse effect on long-term safety of a geological disposal system. In addition, during the emplacement of waste package with buffer material, it is required to limit amount of groundwater inflow to a certain level by grouting. Therefore, it is necessary to develop new grout materials with penetrability for smaller fractures. We have developed new grout materials, which have better penetrability and are environmentally more friendly than exinting cementitious grout materials since FY 2007. This sequel report shows the most appropriate composition and the penetration characteristic of new grout materials to be suitable for the experiment based on the result of indoor test carried out after FY 2008.

Evaluation of fracture toughness of fine-grained isotropic graphites for HTGR

Yamada, Teruaki*; Matsushima, Yuki*; Kuroda, Masatoshi*; Sumita, Junya; Shibata, Taiju; Fujita, Ichiro*; Sawa, Kazuhiro

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 6 Pages, 2012/10

In order to investigate the effects of the experimental methodology and the notch angle on the fracture toughness of the fine-grained isotropic nuclear graphites IG-110 and IG-430, the three-point-bending test, which has been recently proposed as the methodology to evaluate the fracture toughness of graphite for high temperature gas-cooled reactors (HTGRs), was performed using two types of the specimens with different notch angles. The results obtained in this study could be summarized as follows: (1) The values of the fracture toughness of IG-110 and IG-430 measured in this study were 0.890 (MPam) and 1.031 (MPam), respectively. It was also found that the value of the fracture toughness of IG-110 was nearly equal to or smaller than the values obtained by the other method reported previously. (2) The values of the fracture toughness of the fine-grained isotropic graphites were not affected between the notch angles introduced by the incisive razor blade. (3) The ratio of the tensile strengths of IG-110 and IG-430 was estimated from Griffith Theory using the experimental data obtained in this study. The estimated strength ratio was in good agreement with the strength ratio obtained from the supplier's data.

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2010

Sumiya, Shuichi; Watanabe, Hitoshi; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Hiyama, Yoshinori; Yoshii, Hideki*; Kikuchi, Masaaki*; Otani, Kazunori*; Goto, Ichiro*

JAEA-Review 2012-006, 114 Pages, 2012/03

JAEA-Review-2012-006.pdf:3.36MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, and the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and byelaw of Ibaraki prefecture), this report describes the effluent control results of liquid waste discharged from the JAEA's Nuclear Fuel Cycle Engineering Laboratories in the fiscal year 2010, from 1st April 2010 to 31st March 2011. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other facilities were much lower than the authorized limits of the above regulations.

Irradiation-induced dimensional change and fracture behavior of C/C composites for VHTR application

Sumita, Junya; Shibata, Taiju; Sawa, Kazuhiro; Fujita, Ichiro; Ohashi, Jun*; Takizawa, Kentaro*; Kim, W.*; Park, J.*

Ceramic Materials for Energy Applications; Ceramic Engineering and Science Proceedings, Vol.32, No.9, p.1 - 12, 2011/11

https://doi.org/10.1002/9781118095386.ch1

Since the temperature condition in Very High Temperature Reactor (VHTR), one of the Generation-IV reactor systems, is severe, the application of heat-resistant carbon fiber reinforced carbon matrix composite (C/C composite) for control rod elements is one of the important subjects for the VHTR development. JAEA focuses on the application of two-dimensional (2D-) C/C composites for control rod. The 2D-C/C composite has an anisotropy in properties for parallel and perpendicular to lamina directions. Irradiation effects of the 2D-C/C composite also show anisotropic behavior. It is hence important to consider the anisotropy in control rod design. To investigate the irradiation effects of the 2D-C/C composite on properties, irradiation test and post irradiation examination (PIE) were carried out and the irradiation effects were evaluated for the both directions. Since the C/C composite is composed of fibers and matrix, this geometry should be considered to evaluate the crack propagation in the composite. To assess the fracture behavior with crack propagation, bending test was carried out assuming a crack in the control rod and cracks in specimens were observed. This paper describes the irradiation effects of the 2D-C/C composite based on the PIE results considering the anisotropy. The evaluation results on equivalent fracture toughness and fracture mechanism are also discussed.

Study on fracture behavior of 2D-C/C composite for application to control rod of very high temperature reactor

Sumita, Junya; Fujita, Ichiro; Shibata, Taiju; Makita, Taiyo*; Takagi, Takashi*; Kunimoto, Eiji*; Sawa, Kazuhiro; Kim, W. J.*; Park, J. Y.*

IOP Conference Series; Materials Science and Engineering, 18(16), p.162010_1 - 162010_4, 2011/09

https://doi.org/10.1088/1757-899X/18/16/162010

For control rod element of Very High Temperature Reactor, carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials for its high strength and thermal stability. Since the crack propagation of the C/C composite is complicated, the fracture behavior is one of the most important subjects of the design methodology for the control rod with the C/C composite. In this study, in order to assess the Mode-II fracture behavior of the two-dimensional (2D-) C/C composite which has the layer structure of laminas composed of fibers and matrix, bending test with oxidized end notched flexure specimen was carried out. The interlaminar fracture toughness decreased with increasing the oxidation for the 2D-C/C composite. The oxidized matrix causes reduction of fracture toughness and the reduction ratio is dependent on the matrix type. The crack initiates at the boundary between fiber bundles and grows along them without breaking the fiber. It is thought that the cracks which were initiated at the interface between matrix and fiber were gathered into the voids in the boundary between fiber bundles, and then the crack grows up in the matrix.