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Journal Articles

Investigation of high-temperature chemical interaction of calcium silicate insulation and cesium hydroxide

Rizaal, M.; Nakajima, Kunihisa; Saito, Takumi*; Osaka, Masahiko; Okamoto, Koji*

Journal of Nuclear Science and Technology, 57(9), p.1062 - 1073, 2020/09

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

The interaction of cesium hydroxide and a calcium silicate insulation material was experimentally investigated at high temperature conditions. A thermogravimetry equipped with differential thermal analysis was used to analyze thermal events in the samples of mixed calcium silicate and cesium hydroxide under Ar-5%H$$_{2}$$ and Ar-4%H$$_{2}$$-20%H$$_{2}$$0 with maximum temperature of 1100$$^{circ}$$C. Prior being mixed with cesium hydroxide, a part of calcium silicate was pretreated at high temperature to evaluate the effect of possible structural changes of this material due to a preceding thermal history and also the sake of thermodynamic evaluation to those available ones. Based upon the initial condition (preliminary heat treatment) of calcium silicate, it was found that if the original material consisted of xonotlite (Ca$$_{6}$$Si$$_{6}$$0$$_{17}$$(0H)$$_{2}$$), the endothermic reaction with cesium hydroxide occurred over the temperature range 575-730$$^{circ}$$C meanwhile if the crystal phase of original material was changed to wollastonite (CaSi0$$_{3}$$), the interaction occurred over temperature range 700-1100$$^{circ}$$C. Furthermore, the X-ray diffraction analyses have indicated on both type of pretreated calsils that regardless of Ar-5%H$$_{2}$$ and Ar-4%H$$_{2}$$-20%H$$_{2}$$0 atmosphere, cesium aluminum silicate, CsAlSi0$$_{4}$$ was formed with aluminum in the samples as an impurity or adduct.

Journal Articles

Thermal-hydraulic analysis of the LBE spallation target head in JAEA

Wan, T.; Obayashi, Hironari; Sasa, Toshinobu

Nuclear Technology, 205(1-2), p.188 - 199, 2019/01

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

Journal Articles

Considerations on phenomena scaling for BEPU

Nakamura, Hideo

Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 8 Pages, 2018/00

no abstracts in English

Journal Articles

Study on the thermal-hydraulic of TEF-T LBE spallation target in JAEA

Wan, T.; Obayashi, Hironari; Sasa, Toshinobu

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 13 Pages, 2017/09

JAEA Reports

Thermal design study of lead-bismuth cooled accelerator driven system, 1; Study on thermal hydraulic behavior under normal operation condition

Akimoto, Hajime; Sugawara, Takanori

JAEA-Data/Code 2016-008, 87 Pages, 2016/09

JAEA-Data-Code-2016-008.pdf:15.62MB

Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.

Journal Articles

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, 1; Thermal mixing behavior of helium gas in HTTR

Tochio, Daisuke; Fujimoto, Nozomu

Journal of Nuclear Science and Technology, 53(3), p.425 - 431, 2016/03

 Times Cited Count:1 Percentile:82.26(Nuclear Science & Technology)

The future HTGR is now designed in JAEA. The reactor has many merging points of helium gas with different temperature. It is needed to clear the mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the HTTR due to lack of mixing of helium gas in the primary cooling system. Now the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal-hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the mixing behavior of helium gas. As the result, it was confirmed that the mixing behavior of helium gas in the primary cooling system is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.

Journal Articles

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*

Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09

 Times Cited Count:8 Percentile:31.87(Nuclear Science & Technology)

A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.

JAEA Reports

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03

JAEA-Data-Code-2014-031.pdf:37.23MB

A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

Journal Articles

Development of control technology for the HTTR hydrogen production system

Nishihara, Tetsuo; Inagaki, Yoshiyuki

Nuclear Technology, 153(1), p.100 - 106, 2006/01

 Times Cited Count:8 Percentile:46.1(Nuclear Science & Technology)

Japan Atomic Energy Research Institute (JAERI) has performed the research and development of hydrogen production using the high temperature engineering test reactor (HTTR). One of the key issues for the HTTR hydrogen production system is the development of control technology for stable operation. A thermal load absorber concept using a steam generator installed downstream of a reformer is proposed to mitigate a variation of helium temperature. Thermal hydraulic analyses for the start up operation and the suspension of feed gas supply to the reformer are carried out. These results show that a large variation of the reformer outlet helium temperature takes place due to a change of the feed gas flow rate. However the steam generator can mitigate the variation of helium temperature. It is clarified that the HTTR can continue normal operation independently of the feed gas flow rate.

Journal Articles

Large-scale direct simulation of two-phase flow structure around a spacer in a tight-lattice nuclear fuel bundle

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime

Computational Fluid Dynamics 2004, p.649 - 654, 2006/00

no abstracts in English

Journal Articles

Design study around beam window of ADS

Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Kikuchi, Kenji; Kurata, Yuji; Sasa, Toshinobu; Umeno, Makoto*; Nishihara, Kenji; Saito, Shigeru; Mizumoto, Motoharu; Takano, Hideki*; et al.

Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.325 - 334, 2005/11

The Japan Atomic Energy Research Institute (JAERI) is conducting the research and development (R&D) on the Accelerator-Driven Subcritical System (ADS) for the effective transmutation of minor actinides (MAs). The ADS proposed by JAERI is the 800 MWth, Pb-Bi cooled, tank-type subcritical reactor loaded with (MA+Pu) nitride fuel. The Pb-Bi is also used as the spallation target. In this study, the feasibility of the ADS was discussed with putting the focus on the design around the beam window. The partition wall was placed between the target region and the ductless-type fuel assemblies to keep the good cooling performance for the hot-spot fuel pin. The flow control nozzle was installed to cool the beam window effectively. The thermal-hydraulic analysis showed that the maximum temperature at the outer surface of the beam window could be repressed below 500 $$^{circ}$$C even in the case of the maximum beam power of 30 MW. The stress caused by the external pressure and the temperature distribution of the beam window was also below the allowable limit.

JAEA Reports

Benchmark analysis of KRITZ-2 critical experiments

Okumura, Keisuke; Kawasaki, Kenji*; Mori, Takamasa

JAERI-Research 2005-018, 64 Pages, 2005/08

JAERI-Research-2005-018.pdf:3.26MB

In the KRITZ-2 critical experiments, criticality and pin power distributions were measured at room temperature and high temperature (about 245 degree C) for three different cores loading slightly enriched UO$$_{2}$$ or MOX fuels. For nuclear data testing, benchmark analysis was carried out with a continuous-energy Monte Carlo code MVP and its four nuclear data libraries based on JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI.8. As a result, fairly good agreements with the experimental data were obtained with any libraries for the pin power distributions. However, the JENDL-3.3 and ENDF/B-VI.8 give under-prediction of criticality and too negative isothermal temperature coefficients for slightly enriched UO$$_{2}$$ cores, while the older nuclear data JENDL-3.2 and JEF-2.2 give rather good agreements with the experimental data. From the detailed study with an infinite unit cell model, it was found that the differences among the libraries are mainly due to the different fission cross section of U-235 in the energy rage below 1.0 eV.

Journal Articles

Temperature evaluation of core components of HTGR at depressurization accident considering annealing recovery on thermal conductivity of graphite

Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Hanawa, Satoshi; Iyoku, Tatsuo; Ishihara, Masahiro

Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4822 - 4828, 2005/08

Graphite materials are used for structural components in High Temperature Gas-Cooled Reactor (HTGR) core because of their excellent thermo/mechanical properties. Thermal conductivity of graphite components is reduced by neutron irradiation in reactor operation. The reduced conductivity is expected to be recovered by thermal annealing effect when irradiated graphite component is heated above irradiated temperature. In the present study, temperature analyses considering the annealing effect of the HTGR core at a depressurization accident were carried out and influence of annealing effect on maximum fuel temperature was investigated. The analyses show that the annealing effect can reduce the fuel temperature about 100$$^{circ}$$C at the maximum, and it is possible to evaluate the maximum fuel temperature more appropriately. It was also shown that the core-temperature of High Temperature Engineering Test Reactor (HTTR) at the safety demonstration tests can be analyzed with the developed evaluation method considering annealing effect.

Journal Articles

Natural convection heat transfer of high temperature gas in an annulus between two vertical concentric cylinders

Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki

Heat Transfer-Asian Research, 34(5), p.293 - 308, 2005/07

Water cooling panels have been adopted as the vessel cooling system of the HTTR to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate the heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with thermal radiation heat transfer in an annulus between two vertical concentric cylinders with the inner cylinder heated and the outer cylinder cooled. In the present experiments, Rayleigh number based on the height of the annulus ranged from 2.0$$times$$10$$^{7}$$ to 5.4$$times$$10$$^{7}$$ for helium gas and from 1.2$$times$$10$$^{9}$$ to 3.5$$times$$10$$^{9}$$ for nitrogen gas. The numerical results were in good agreement with the experimental ones regarding the surface temperatures of the heating and cooling walls. As a result of the experiments and the numerical analyses, the heat transfer coefficient of natural convection coupled with thermal radiation was obtained.

Journal Articles

Thermal-hydraulic responses during PWR pressure vessel upper head small break LOCA based on LSTF experiment and analysis

Takeda, Takeshi; Asaka, Hideaki; Suzuki, Mitsuhiro; Nakamura, Hideo

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05

no abstracts in English

Journal Articles

Numerical analysis of three-dimensional two-phase flow behavior in a fuel assembly

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime

WIT Transactions on Engineering Sciences, Vol.50, p.183 - 192, 2005/00

no abstracts in English

JAEA Reports

Preliminary investigation of annealing effect on thermal conductivity of graphite and investigation of annealing test method (Contract research)

Sumita, Junya; Nakano, Masaaki*; Tsuji, Nobumasa*; Shibata, Taiju; Ishihara, Masahiro

JAERI-Tech 2004-055, 25 Pages, 2004/08

JAERI-Tech-2004-055.pdf:4.25MB

Neutron irradiation remarkably reduces the thermal conductivity of graphite, and the reduced thermal conductivity is recovered by annealing effect if the graphite is heated above the irradiation temperature. Therefore, it is expected that the reduced thermal conductivity of graphite components in the HTGR could be recovered by the annealing effect in accidents, such as a depressurization accident. Then, an analytical investigation of the annealing effect on thermal performance of a HTGR core was carried. The analysis showed that the annealing effect reduces the maximum fuel temperature about 70$$^{circ}$$C, and it is important to introduce the annealing effect appropriately in the temperature analysis of the core components and reactor internals. In addition, an annealing test method was investigated to evaluate the effect quantitatively, and the test plan was made.

Journal Articles

Study on natural convection heat transfer of high temperature gas in a vertical annular space of a double coaxial cylinder

Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki

Nippon Kikai Gakkai Rombunshu, B, 70(694), p.1518 - 1525, 2004/06

no abstracts in English

Journal Articles

Development of plant dynamics analytical code named Conan-GTHTR for the Gas Turbine High Temperature Gas-cooled Reactor, 1; Code validation by Use of the experimental data of HTTR

Takamatsu, Kuniyoshi; Katanishi, Shoji; Nakagawa, Shigeaki; Kunitomi, Kazuhiko

Nippon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.76 - 87, 2004/03

The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in Higt Temperature Gas-Cooled Reactor systems.

Journal Articles

Effects of process parameters of the IS process on total thermal efficiency to produce hydrogen from water

Kasahara, Seiji; Hwang, G.*; Nakajima, Hayato; Choi, H.*; Onuki, Kaoru; Nomura, Mikihiro

Journal of Chemical Engineering of Japan, 36(7), p.887 - 899, 2003/07

 Times Cited Count:55 Percentile:12.24(Engineering, Chemical)

Thermal efficiency of the IS thermochemical hydrogen production process was evaluated. Sensitivities of operation conditions (HI conversion ratio, pressure and reflux ratio at HI distillation and concentration of HI after EED) and nonidealities of the process (electric energy loss in EED, loss at heat exchangers and loss of waste heat recovery as electricity) were investigated. Concentration of HI after EED had the most significant effect of 13.3 % on thermal efficiency in operation conditions. Nonidealities had importance on thermal efficiency. Thermal efficiency was 56.8 % with optimized operation conditions and no nonidealities.

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