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JAEA Reports

Improvements of the Control Rod Material for Long Lifetime by the High Power Millimeter-wave Ceramics Sintering Method

Idehara, Toshitaka*; Mitsudo, Seitaro*; Hoshizuki, Hisanori*; Ogawa, Isamu*; Shibahara, Itaru; Nishi, Hiroshi; Kitano, Akihiro; Ishibashi, Junichi

JNC TY4400 2003-005, 106 Pages, 2003/03

JNC-TY4400-2003-005.pdf:9.48MB

Boron Carbide B4C pellet is an important part of control rod used to control the reactivity of nuclear reactors. B4C pellet put in a nuclear reactor suffers heavy radiation damage and deformation, which result in a partial destruction and shorten the lifetime of B4C. It is important to improve the characteristics of the B4C pellets for extension of its lifetime. As the results, if the control rod without the shroud will be available, we can realize much simpler structure. In order to improve, the B4C pellet, which was sintered by the hot-press methods, we have re-sintered it by high power millimeter wave ceramics nano-indentation test. The increase of the plasticity is observed. The same improvement of plasticity was observed in alumina pellets that were sintered by millimeter wave sintering methods. Such results imply that the further improvement is expected, if the B4C pellet is sintered from powder specimen by the high power millimeter-wave sintering method. To simulate a partial destruction of B4C pellet under the thermal stress, preliminary internal heating experiments of B4C pellet are performed by using high power millimeter-wave. At the difference between internal and surface temperatures of 1000C, the partial destructions and small cracks are observed in B4C pellet. Thes may be a kind of model experiment for destruction of B4C pellet irradiated by neutrons.

JAEA Reports

The Basic examination on the development of long-life control rod for prototype reactor Monju; Evaluation of the porous-plug properties in liquid sodium

Yoshida, Eiichi; Sakurai, Tadashi*; Shibahara, Itaru

JNC TN9400 2003-050, 77 Pages, 2003/03

JNC-TN9400-2003-050.pdf:5.93MB

Sodium-bond double porous-plug is one of the candidates in development of the long-life life control rod for prototype reactor MONJU. Gas hold pressure and sodium penetration pressure of the porous-plug in liquid sodium are an important factor which determines the filling up of sodium into porous-plug and the gas plenum length in control rod pin. In this study, the fundamental characteristic of porous-plug was clarified by experiment in high temperature sodium. Moreover, the characteristic equation of design for a porous-plug was evaluated based on the obtained data.

Journal Articles

B$$_{4}$$C ceramics sintering using 24 GHz gyrotron radiation

Hoshizuki, Hisanori*; Kuroda, Tsutomu*; Mitsudo, Seitaro*; Idehara, Toshitaka*; Glyavin, M.*; Kitano, Akihiro; Ishibashi, Junichi; Nishi, Hiroshi; Shibahara, Itaru

En Sekigai Ryoiki Kaihatsu Kenkyu, 4, p.179 - 185, 2003/00

None

Journal Articles

Millimeter wave sintering of B$$_{4}$$C by using a compact gyrotron system

Hoshizuki, Hisanori*; Mitsudo, Seitaro*; Glyavin, M.*; Eremeev, A.*; *; *; Kitano, Akihiro; Ishibashi, Junichi; Nishi, Hiroshi; Shibahara, Itaru; et al.

Proceedings of 28th International Conference on Infrared and Millimeter Waves (IRMMW 2003), P. 1005, 2003/00

None

Journal Articles

Feasibility Study on Application of Ductless Fuel Assembly to FBR

Shibahara, I.; Shibahara, Itaru

Proceedings of 4th International Conference on Nuclear Engineering (ICONE-4), P. 219, 1996/00

Journal Articles

Effects of neutron irradiation on creep properties of FBR grade 316 stainless steel

Aoto, Kazumi; Abe, Yasuhiro; Shibahara, Itaru; Wada, Yusaku

IAEA-TECDOC-817, p.27 - 37, 1995/00

None

Journal Articles

Present status of radiation damage mechanism study on reactor pressure vessels based on irradiation correlatin concept

Ishino, Shiori*; Sekimura, Naoto*; Suzuki, Masahide; *; *; Shibahara, Itaru*

Nihon Genshiryoku Gakkai-Shi, 36(5), p.396 - 404, 1994/00

 Times Cited Count:1 Percentile:18.18(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Development of Advanced Austenitic Stainless Steel for Fast Reactor Core Material

Shikakura, Sakae; Ukai, Shigeharu; Sato, Yoshinori; Harada, Makoto; Koyama, Shinichi; ; Nomura, Shigeo; Shibahara, Itaru

Nihon Genshiryoku Gakkai-Shi, 36(5), p.441 - 455, 1994/00

None

Journal Articles

Irradiation Performance of modified 316 Stainless Steel for Monju Fuel

Shibahara, Itaru; Ukai, Shigeharu; Onose, Shoji; Shikakura, Sakae

Journal of Nuclear Materials, 204(2), p.131 - 140, 1993/09

 Times Cited Count:39 Percentile:94.74(Materials Science, Multidisciplinary)

None

Journal Articles

Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR

Tsukada, Takashi; Jitsukawa, Shiro; Shiba, Kiyoyuki; Sato, Yoshinori*; Shibahara, Itaru*; Nakajima, Hajime

Journal of Nuclear Materials, 207, p.159 - 168, 1993/00

 Times Cited Count:6 Percentile:56.97(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Evaluation of driver fuel performance in the Joyo Mk-II core

Asaka, Takeo; ; ; ; Shibahara, Itaru; Shikakura, Sakae

Journal of Nuclear Materials, 204, p.102 - 108, 1993/00

 Times Cited Count:4 Percentile:45.58(Materials Science, Multidisciplinary)

None

JAEA Reports

Electrochemical corrosion behavior of stainless steel irradiated in FBR

Shiba, Kiyoyuki; Tsukada, Takashi; Nakajima, Hajime; ; Kitagawa, Isamu; ; Sekino, Hajime; ; ; Itonaga, Fumio; et al.

JAERI-M 92-166, 27 Pages, 1992/11

JAERI-M-92-166.pdf:2.35MB

no abstracts in English

JAEA Reports

Irradiation assisted stress corrosion cracking of stainless steel irradiated in FBR, 1

Tsukada, Takashi; Shiba, Kiyoyuki; Nakajima, Hajime; Usui, Takeshi; Omi, Masao; Goto, Ichiro; ; Nakagawa, Tetsuya; Kawamata, Kazuo; ; et al.

JAERI-M 92-165, 41 Pages, 1992/11

JAERI-M-92-165.pdf:4.99MB

no abstracts in English

JAEA Reports

Integrity evaluations for the 2nd Fugen pressure tube surveillance test

; ; ; ; ; Shibahara, Itaru

PNC TN9410 92-321, 30 Pages, 1992/10

PNC-TN9410-92-321.pdf:0.67MB

Integrity evaluations have been performed for the 2nd Fugen pressure tube test (8 years irradiation, 5.6 $$times$$ 10$$^{21}$$n/cm$$^{2}$$ (E$$>$$1Mev)). Test items mainly consist of tensile test, bending test, corrosion test and hydrogen analysis. It has become clear using these data that the pressure tube material has maintained its integrity during the irradiation by the integrity assessment on both tensile and fracture toughness properties. Besides, both thickness loss by corrosion and absorbed hydrogen content were lower than those of design values.

Journal Articles

Evaluation of irradiation assited stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR

Sato, Yoshinori; Shibahara, Itaru; *; *

Journal of Nuclear Materials, 0 Pages, 1992/00

None

Journal Articles

Stress corrosion cracking tests of stainless steel irradiated in FBR, 1; Interim report of JAERI/PNC cooperative research

Tsukada, Takashi; Shiba, Kiyoyuki; Nakajima, Hajime; Sato, Yoshinori*; Shibahara, Itaru*

PNC-TN9410 92-295, 67 Pages, 1992/00

no abstracts in English

JAEA Reports

Irradiation behavior of the B$$_{4}$$C neutron absorber materials; Results of the JOYO AMIR irradiation experiments

Kimura, Yoshio; Kaito, Takeji; Onose, Shoji; Sato, Yoshinori; Shibahara, Itaru

PNC TN9410 91-171, 17 Pages, 1991/04

PNC-TN9410-91-171.pdf:0.66MB

The B$$_{4}$$C neutron absorber material has great advantage of high neutron absorbing capability, although the He release and the swelling limit the life time of control rod. So it is important for the life time extension and reliability improvement to investigate the irradiation behavior of the B$$_{4}$$C. This report discuss the important parameters which may control the absorber pin behavior based on the irradiation test results obtained recently by the JOYO AMIR experiments, and re-evaluate the current technical issues and the way of improvements. The He release from B$$_{4}$$C pellet increase rapidly after a burnup of 100$$times$$10$$^{20}$$ cap/cm$$^{3}$$ and exceed the value predicted by the equation which has been applied to the MONJU design. The data is still required when we adopt the seal type absorber pin to use at high burnup. The He release found to depend on the $$^{10}$$B enrichment as well as the grain size. The He release could be reduced by the optimization of these parameters although the swelling behavior should also be considered since the swelling generally increases with decreasing the He release. The absorber cladding mechanical interaction (ACMI) is accelerated due to pellet cracking and relocation and essentially limits the life time of control rod. The relocation becomes generally extensive as the pellet-to-cladding gap is increased, but this behavior consider to be stochastic and difficult to make quantitative prediction. Improvement could be attained by using a shroud type absorber pin. Reduction of thermal conductivity during irradiation is considered to be caused by micro-defects including the micro-cracking. The degradation may be improved by introducing B$$_{4}$$C cermet.

JAEA Reports

None

*; Kukita, Shimpei; Yoshinaka, Kazuyuki; *; *; Shibahara, Itaru*

PNC TN9410 89-190, 48 Pages, 1989/11

PNC-TN9410-89-190.pdf:3.48MB

None

JAEA Reports

None

Shikakura, Sakae*; Matsushima, Hideya*; Shibahara, Itaru*; *; *

PNC TN9420 89-004, 124 Pages, 1989/10

PNC-TN9420-89-004.pdf:2.6MB

None

JAEA Reports

None

*; Ukai, Shigeharu; Unno, Ichiro; ; Shibahara, Itaru*; Enokido, Yuji*

PNC TN9410 88-206, 71 Pages, 1988/12

PNC-TN9410-88-206.pdf:4.01MB

None

48 (Records 1-20 displayed on this page)