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Journal Articles

Subchannel analysis of thermal-hydraulics in a fuel assembly with inner duct structure of a sodium-cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Journal of Nuclear Engineering and Radiation Science, 5(2), p.021001_1 - 021001_12, 2019/04

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. In this study, thermal-hydraulics in FAIDUS was investigated with the in-house subchannel analysis code named ASFRE. Before the application to FAIDUS, applicability of ASFRE to FAs was confirmed through the numerical simulations for the experiments of simulated FA. Through the comparisons between the numerical results of thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct, it was indicated that significant asymmetric temperature distribution would not occur in FAIDUS at both high and low flow rate conditions.

Journal Articles

Thermal-hydraulic analysis of fuel assembly with inner duct structure of an advanced loop-type sodium-cooled fast reactor using ASFRE code

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07

In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.

Journal Articles

Development of thermal hydraulics analysis code ASFRE for fuel assembly of sodium-cooled fast reactor; Modification of distributed resistance model and validation analysis

Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.

Journal Articles

Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

The subchannel analysis code NASCA was applied to critical power prediction of 37-rod tight-lattice bundle experiments which JAERI has been carrying out to confirm the thermal-hydraulic feasibility of the RMWR. The NASCA can yield good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy of critical power deteriorated in case of the gap width of 1.0 mm. Predicted BT positions agree with the experimental results. Models in the code will be improved to consider the effect of the gap width based on further studies in the future.

JAEA Reports

Rod displacement measurements by X-ray CT and its impact on thermal-hydraulics in tight-lattice rod bundle (Joint research)

Mitsutake, Toru*; Katsuyama, Kozo*; Misawa, Takeharu; Nagamine, Tsuyoshi*; Kureta, Masatoshi*; Matsumoto, Shinichiro*; Akimoto, Hajime

JAERI-Tech 2005-034, 55 Pages, 2005/06

JAERI-Tech-2005-034.pdf:7.76MB

In tight-lattice bundles with about 1mm gap between rods, a rod displacement might affect thermal-hydraulic characteristics. The inside-structure observation of the simulated seven-rod bundle of RMWR was made with the high-energy X-ray CT of JNC. The CT view assured that the rod position was almost the same as expected by design. In the heat transfer experiments, all thermocouples on the center rod showed almost simultaneous BT-induced temperature increase and on the same axial heights showed quite similar time-variation behaviors in the vapor cooling heat transfer regime. It showed that the effect of the geometrical asymmetry was small on the BT characteristics. The calculated critical power by subchannel analysis with the input of the CT measured rod position was smaller by about 5% than that with the designed rod position. It concluded that the error in the calculated critical power was attributable not to the asymmetry in the rod position, but to the models in the subchannel analysis code.

Journal Articles

Rod displacement effect on thermal-hydraulic behaviour in tight-lattice bundle based on X-ray CT measurement

Mitsutake, Toru*; Akimoto, Hajime; Misawa, Takeharu; Kureta, Masatoshi*; Katsuyama, Kozo*; Nagamine, Tsuyoshi*; Matsumoto, Shinichiro*

Proceedings of 4th World Congress on Industrial Process Tomography, Vol.1, p.348 - 353, 2005/00

An inside-structure observation of a tight-lattice 7-rod bundle was made, using the high-energy X-ray computer tomography(CT) apparatus. The two-dimensional configurations of all rods were obtained at seventy-six axial height positions over the whole length of the bundle. The measured results of the rod positions showed small rod position displacements, about 0.5 millimeters at maximum, from the lattice positions. Based on these measured rod position displacement results, the flow area, equivalent hydraulic diameter, rod-rod clearance, and rod-shroud clearance were calculated. The effect of rod position displacement on critical power was estimated by a sub-channel analysis. The result showed that the rod position displacement effect has only a small effect on critical power calculations. The calculated critical power still overestimated the measured value.

Journal Articles

Subchannel analysis of CHF experiments for tight-lattice core

Nakatsuka, Toru; Tamai, Hidesada; Kureta, Masatoshi; Okubo, Tsutomu; Akimoto, Hajime; Iwamura, Takamichi

Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 6 Pages, 2003/09

It is important to evaluate thermal margin of the tight lattice core in the Reduced-Moderation Water reactor (RMWR). In the present study, to assess the applicability of subchannel analysis for tight lattice cores, tight lattice CHF experiments were analyzed with COBRA-TF code. For the axial uniform heated rod bundle, the code gives good prediction of critical power for mass velocity of around 500kg/(m$$^{2}$$s), while the code underestimates it for lower mass velocity and overestimates for higher mass velocity. The predicted BT position was outer channels and differed from the measured position. For the axially double-humped heated bundle, the code gives good prediction for mass velocity of around 200kg/(m$$^{2}$$s), and overestimates for higher mass velocity. It turned out that the two-phase multiplier of friction loss have a large influences on the flow distribution among the subchannels. To improve the calculation accuracy, it is required to predict precisely the flow distribution including the prediction of pressure distribution in a tight lattice bundle.

Journal Articles

Study on gas-liquid two-phase flow distribution in a tight-lattice rod bundle

Onuki, Akira; Shibata, Mitsuhiko; Tamai, Hidesada; Akimoto, Hajime; Yamauchi, Toyoaki*; Mizokami, Shinya*

Nippon Konsoryu Gakkai Nenkai Koenkai 2003 Koen Rombunshu, p.35 - 36, 2003/07

Analytical evaluation of maximum critical power by so-called subchannnel code is indispensable for design of reduced moderation water reactor. In this study, two-phase flow distribution in a tight-lattice rod bundle is investigated using 19-rod bundle experimental rig and subchannnel analysis code NASCA. The flow distribution was measured under so-called churn flow regime and the predictive capability of NASCA was assessed. NASCA can predict the flow distribution qualitatively depending on local pressure drop. Quantitative prediction is also reasonable for liquid phase but the gas phase distribution was underestimated. Void-drift model has a dominant contribution and we should improve the model for the tight-lattice rod bundle.

JAEA Reports

CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu; Kureta, Masatoshi; Yoshida, Hiroyuki; Ishikawa, Nobuyuki; Sato, Takashi; Watanabe, Hironori; Okubo, Tsutomu; Iwamura, Takamichi; et al.

JAERI-Research 2002-018, 37 Pages, 2002/10

JAERI-Research-2002-018.pdf:2.62MB

no abstracts in English

Journal Articles

CHF experiments under steady-state and transient conditions for tight lattice core with non-uniform axial power distribution

Iwamura, Takamichi; Watanabe, Hironori; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(5), p.413 - 424, 1993/05

 Times Cited Count:1 Percentile:80.42(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Application of subchannel code to DNB analysis of HCLWR

Iwamura, Takamichi; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio

Subchannel Analysis in Nuclear Reactors, p.281 - 301, 1992/00

no abstracts in English

Journal Articles

Evaluation of DNBR under operational and accident conditions for double-flat-core type HCLWR

Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio; *; *

Journal of Nuclear Science and Technology, 28(1), p.45 - 58, 1991/01

no abstracts in English

JAEA Reports

Analytical study on the effects of ralling on the flow in the core of marine reactor

Kusunoki, Tsuyoshi; *; ; *

JAERI-M 90-181, 57 Pages, 1990/10

JAERI-M-90-181.pdf:1.04MB

no abstracts in English

JAEA Reports

Critical heat flux experiments for high conversion light water reactor, 3; Heated length: 0.5$$sim$$1.0m, P/D: 1.126$$sim$$1.2, number of rods: 4$$sim$$7

Iwamura, Takamichi; Okubo, Tsutomu; *; ; Murao, Yoshio

JAERI-M 90-044, 158 Pages, 1990/03

JAERI-M-90-044.pdf:2.81MB

no abstracts in English

JAEA Reports

DNBR analyses under steady-state and accident conditions for a double-flat-core high conversion light water reactor

Iwamura, Takamichi; *; Okubo, Tsutomu; ; Murao, Yoshio

JAERI-M 90-043, 70 Pages, 1990/03

JAERI-M-90-043.pdf:1.77MB

no abstracts in English

JAEA Reports

Effect of two-phase mixing coefficient on subchannel analysis and CHF prediction

Iwamura, Takamichi; Okubo, Tsutomu; *; ; Murao, Yoshio

JAERI-M 89-127, 47 Pages, 1989/09

JAERI-M-89-127.pdf:1.18MB

no abstracts in English

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