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Nakayama, Shinsuke; Iwamoto, Osamu; Kimura, Atsushi
EPJ Web of Conferences, 294, p.07001_1 - 07001_6, 2024/04
Graphite is a candidate of moderator in innovative nuclear reactors such as molten salt reactors. Scattering of thermal neutrons by the moderator material has a significant impact on the reactor core design. To contribute to the development of innovative nuclear reactors, an evaluation method of thermal neutron scattering law for reactor grade graphite was studied. The inelastic scattering component due to lattice vibration was evaluated based on the phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The coherent elastic scattering component due to crystal structure was evaluated based on neutron transmission and scattering experiments recently performed in the J-PARC/MLF facility. In comparison with the neutron transmission experiments, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important. Based on the above methods, thermal neutron scattering law data for reactor-grade graphite at room temperature were evaluated.
Klotz, S.*; Baptiste, B.*; Hattori, Takanori; Feng, S. M.*; Jin, Ch.*; B
neut, K.*; Guigner, J. M.*; Est
ve, I.*
Carbon, 185, p.491 - 500, 2021/11
Times Cited Count：1 Percentile：6.01(Chemistry, Physical)Carbon disulphide (CS) is one of the simplest molecular systems made of double covalent bonds. Under high pressure, the molecular structure is expected to break up to form extended crystalline or polymeric solids. Here we show that by compression at 300 K to approximately 
10 GPa using large-volume high pressure techniques, an instantaneous reaction leads to a mixture of pure sulphur and a well-defined compound with stoichiometry close to CS which can be recovered to ambient pressure. We present neutron and X-ray diffraction as well as Raman data which show that this material consists of sulphur bonded to sp
graphite layers of nanometric dimensions. The compound is a semiconductor with a gap of 45 meV, as revealed by temperature dependent resistivity measurements, and annealing at temperatures above 200
C allow to reduce its sulphur content up to C
S. Its structural and electronic properties are fundamentally different to "Bridgman black" reported from previous high pressure experiments on CS.
Hamamoto, Shimpei; Ohashi, Hirofumi; Iigaki, Kazuhiko; Shimazaki, Yosuke; Ono, Masato; Shimizu, Atsushi; Ishitsuka, Etsuo
Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), 6 Pages, 2021/10
Since the HTGR has a large amount of graphite material in the core, it is necessary to assume an accident in which the reactor pressure boundary is damaged and air flows into the core. It is important to state that at the time of this accident, graphite does not burn and the accident does not develop due to the heat of oxidation reaction. Therefore, in this study, in order to evaluate the combustibility of graphite materials, we propose a method to compare the calorific value and heat removal amount of the material. When calculating the calorific value, the structural material of HTTR, a high-temperature gas reactor in Japan, was used as a reference. The amount of air in contact with the structural material is a value determined from the chimney effect. The amount of heat release is the sum of convection and radiation. As a result of comparing the heat generation amount with the heat removal amount, it was shown that the heat release amount was always larger than the heat generation amount. This result shows that the graphite material does not depend on the state at the time of the air inflow accident, the temperature decreases and does not burn. It is important to clearly explain the non-flammability of graphite materials when deciding how to deal with severe accidents in HTGRs. This quantitative evaluation method based on a simple theory is considered useful.
Np using graphite thermal columnNakamura, Shoji; Endo, Shunsuke; Kimura, Atsushi; Shibahara, Yuji*
KURNS Progress Report 2020, P. 94, 2021/08
The present study selected Np among radioactive nuclides and aimed to converge a contradiction between reported thermal-neutron capture cross sections. Neutron irradiation was carried out using the graphite thermal column equipped with the Kyoto University Research Reactor. A solution equivalent to 950 Bq order of radioactivity was pipetted out of a Np standard solution and dropped onto a fiber filter, which was then dried with an infrared lamp to prepare a Np sample. The Np sample was quantified using 312-keV gamma ray emitted from 
Pa in a radiation equilibrium with Np. To monitor a thermal-neutron flux component at an irradiation position, the Np sample was irradiated together with several stable nuclides as neutron flux monitors: 
Sc, 
Co, 
Mo, 
Ta and 
Au. The reaction rate of Np was obtained from gamma-ray yields given by 
Np and Pa, and then the thermal-neutron capture cross section of Np was derived.
Fukaya, Yuji; Goto, Minoru; Nakagawa, Shigeaki; Nakajima, Kunihiro*; Takahashi, Kazuki*; Sakon, Atsushi*; Sano, Tadafumi*; Hashimoto, Kengo*
EPJ Web of Conferences, 247, p.09017_1 - 09017_8, 2021/02
The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce a generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment to observe subcriticality. To achieve the objectives, the reactor core of graphite-moderation system named B7/4"G2/8"p8EUNU+3/8"p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core is composed of the fuel assembly, driver fuel assembly, graphite reflector, and polyethylene reflector. The fuel assembly is composed of enriched uranium plate, natural uranium plate and graphite plates to realize the average fuel enrichment of HTTR and it's spectrum. However, driver fuel assembly is necessary to achieve the criticality with the small-sized core. The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In this study, the overview of the criticality experiments is reported. The reactor configuration with graphite moderation system is rare case in the KUCA experiments, and this experiment is expected to contribute not only for an HTGR development but also for other types of a reactor in the graphite moderation system such as a molten salt reactor development.
Sakon, Atsushi*; Nakajima, Kunihiro*; Takahashi, Kazuki*; Hohara, Shinya*; Sano, Tadafumi*; Fukaya, Yuji; Hashimoto, Kengo*
EPJ Web of Conferences, 247, p.09009_1 - 09009_8, 2021/02
In graphite-reflected thermal reactors, even a detector placed far from fuel region may detect a certain degree of the correlation amplitude. This is because mean free path of neutrons in graphite is longer than that in water or polyethylene. The objective of this study is experimentally to confirm a high flexibility of neutron detector placement in graphite reflector for reactor noise analysis. The present reactor noise analysis was carried out in a graphite-moderated and -reflected thermal core in Kyoto University Critical Assembly (KUCA). BF
proportional neutron counters (1" dia.) were placed in graphite reflector region, where the counters were separated by about 35cm and 30cm -thick graphite from the core, respectively. At a critical state and subcritical states, time-sequence signal data from these counters were acquired and analyzed by a fast Fourier transform (FFT) analyzer, to obtain power spectral density in frequency domain. The auto-power spectral density obtained from the counters far from the core contained a significant degree of correlated component. A least-squares fit of a familiar formula to the auto-power spectral density data was made to determine the prompt-neutron decay constant. The decay constant was 63.3
14.5 [1/s] in critical state. The decay constant determined from the cross-power spectral density and coherence function data between the two counters also had a consistent value. It is confirmed that reactor noise analysis is possible using a detector placed at about 35cm far from the core, as we expected.
Mineo, Hideaki; Nishihara, Tetsuo; Ohashi, Hirofumi; Goto, Minoru; Sato, Hiroyuki; Takegami, Hiroaki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 62(9), p.504 - 508, 2020/09
High-Temperature Gas-cooled Reactor (HTGR) is one of thermal neutron reactor-type that employs helium gas coolant and graphite moderator. It has excellent inherent safety and can supply high-temperature heat which can be used not only for electric power generation but also for a wide range of application such as hydrogen production. Therefore, HTGR is expected to be an effective technology for reducing greenhouse gases in Japan as well as overseas. In this paper, we will introduce the forefront of technological development that JAEA is working toward the realization of an HTGR system consisting of a high temperature gas reactor and heat utilization facilities such as gas-turbine power generation and hydrogen production.
Shibata, Taiju; Mizuta, Naoki; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; et al.
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10
Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). Oxidation damage on the graphite components in air ingress accident is a crucial issue for the safety point of view. SiC coating on graphite surface is a possible technique to enhance oxidation resistance. However, it is important to confirm the integrity of this material against high temperature and neutron irradiation for the application of the in-core components. JAEA and Japanese graphite companies carried out the R&D to develop the oxidation-resistant graphite. JAEA and INP investigated the irradiation effects on the oxidation-resistant graphite by using a framework of ISTC partner project. This paper describes the results of post irradiation experiment about the neutron irradiated SiC-coated oxidation-resistant graphite. A brand of oxidation-resistant graphite shows excellent performance against oxidation test after the irradiation.
Mizuta, Naoki; Sumita, Junya; Shibata, Taiju; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Sakaba, Nariaki
Tanso Zairyo Kagaku No Shinten; Nihon Gakutsu Shinkokai Dai-117-Iinkai 70-Shunen Kinen-Shi, p.161 - 166, 2018/10
To enhance oxidation resistance of graphite material for in-core components of HTGR, JAEA and four Japanese graphite companies; Toyo Tanso, IBIDEN, Tokai Carbon and Nippon Techno-Carbon, are carrying out for development of oxidation-resistant graphite by CVD-SiC coating. This paper describes the outline of neutron irradiation test about the oxidation-resistant graphite by WWR-K reactor of INP, Kazakhstan through an ISTC partner project. Prior to the irradiation test, the oxidation-resistant graphite by CVD-SiC coating of all specimens showed enough oxidation resistance under un-irradiation condition. The neutron irradiation test was already completed and out-of-pile oxidation test will be carried out at the hot-laboratory of WWR-K.
Shibata, Taiju; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.567 - 571, 2016/11
Graphite are used for the in-core components of HTGR, and it is desirable to enhance oxidation resistance to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies are studying to develop oxidation-resistant graphite. Neutron irradiation test was carried out by WWR-K reactor of INP of Kazakhstan through ISTC partner project. The total irradiation cycles of WWR-K operation was 10 cycles by 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E 
0.18 MeV) for the capsule irradiated at a central irradiation hole was preliminary calculated as 1.2
10
/m
, and for the capsule at a peripheral irradiation hole as 4.210
/m. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory.
Uchida, Yoshiaki*; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio
JAERI-Tech 2005-012, 31 Pages, 2005/03
JAERI-Tech-2005-012.pdf:4.58MB
no abstracts in English
Yagisawa, Hiroshi; Arai, Takashi; Goto, Yoshitaka*
Heisei-16-Nendo Osaka Daigaku Sogo Gijutsu Kenkyukai Hokokushu (CD-ROM), 4 Pages, 2005/03
no abstracts in English
Ishihara, Masahiro
Seramikkusu, 39(10), p.834 - 837, 2004/10
no abstracts in English
Goto, Yoshitaka*; Arai, Takashi; Yagyu, Junichi; Masaki, Kei; Kodama, Kozo; Miya, Naoyuki
Journal of Nuclear Materials, 329-333(1), p.840 - 844, 2004/08
Times Cited Count：15 Percentile：68.16(Materials Science, Multidisciplinary)TEM and Selected Area Diffraction (SAD) were made on nm-structures of redeposition layers on graphite tiles used in the lower-X-point divertor of JT-60. The tiles were used in the 1988 experimental campaign in which 300 divertor discharges and 1500 limiter discharges were made. TEM observations were made at poloidal and/or toroidal sections at two positions on the inboard side of the inner-separatrix strike point. Layer structures in 0-6micron depths were correlated to the last 40-shots in the campaign. Columnar structures corresponded to divertor discharges of additional heating power below 10MW. Lamellar structures were due to limiter discharges or to the higher power divertor discharges. Carbon-Mo, Ti or carbon-Ni, Fe, Cr, Ti codeposition layers were ascribed to disruptive shots. From analyses on poloidal orientation of the column axes and graphene sheets composing columns, the observed columnar structures were ascribed to both low adatom-migration due to the low deposition temperatures and also to self-shadowing effects due to inclined incidence of carbon impurity ions.
Goto, Yoshitaka*
Kaku Yugoro, (11), p.34 - 37, 2004/03
Recent results of erosion and re-deposition studies are reported for graphite tiles from W-shaped divertor region of JT-60U. The tiles were operated in Jun. 1997 - Oct. 1998 periods in which more than 3000 D-D discharge experiments were made with all-carbon walls with 2-times boronizations and inner-private flux pumping. Erosion depth was estimated by using dial gauge, while deposition was measured with SEM. On the outer divertor target, erosion was found dominant, while on the inner target, re-deposition was found to be dominant. No continuous deposition layers on the dome top. Observed in/out asymmetry is attributable to in/out asymmetry of plasma particle conditions in front of the divertor plates. Columnar structures in the redeposition layers corresponded to the lower heat-flux zone, while lamellar structures were found in the overlayers in the higher heat-flux zone near the separatrix strike point.
Katanishi, Shoji; Takei, Masanobu; Nakata, Tetsuo*; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.67 - 75, 2004/03
no abstracts in English
Tomimitsu, Hiroshi; Hasegawa, Yuji*; Aizawa, Kazuya
Physics Letters A, 309(3-4), p.183 - 188, 2003/03
Times Cited Count：0 Percentile：0.01(Physics, Multidisciplinary)no abstracts in English
Oku, Tatsuo*; Baba, Shinichi
Tanso, 2002(202), p.90 - 95, 2002/05
no abstracts in English
Kawasaki, Kozo
Genshiryoku Nenkan 2003-Nen Ban, p.150 - 158, 2002/00
no abstracts in English
Hagiwara, Masaki
Genshiryoku Nenkan 2001/2002-Nen Ban, p.164 - 170, 2001/11
no abstracts in English
