Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 52

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Numerical simulation of two-phase flow in 4$$times$$4 simulated bundle

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06

JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.

Journal Articles

Experimental study on local interfacial parameters in upward air-water bubbly flow in a vertical 6$$times$$6 rod bundle

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 144, p.118696_1 - 118696_19, 2019/12

 Times Cited Count:0 Percentile:100(Thermodynamics)

Journal Articles

Local gas-liquid two-phase flow characteristics in rod bundle geometry

Xiao, Y.*; Shen, X.*; Miwa, Shuichiro*; Sun, Haomin; Hibiki, Takashi*

Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08

In order to develop constitutive equations of two-fluid model in rod bundle flow channels, experiments of adiabatic air-water upward two-phase flow in 6$$times$$6 rod bundle flow channel were performed. Local flow parameters such as void fraction, interfacial area concentration (IAC) and so on were measured by a double-sensor optical probe. The area-averaged void fraction and IAC data were compared with the predictions from a drift-flux model and an IAC correlation.

Journal Articles

Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Ito, Masahiro*

Nuclear Engineering and Design, 317, p.133 - 145, 2017/06

 Times Cited Count:5 Percentile:36.05(Nuclear Science & Technology)

The coupled numerical analysis of mechanical and thermal behaviors was performed for a wire-wrap fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal hydraulics analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that radial distribution of coolant temperatures in a subassembly tended to be flattened as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such temperature distribution was slightly analyzed as a result of the small bowing of the fuel pins due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal hydraulics was also investigated in this study.

Journal Articles

Heat conduction analyses on rewetting front propagation during transients beyond anticipated operational occurrences for BWRs

Yonomoto, Taisuke; Shibamoto, Yasuteru; Satou, Akira; Okagaki, Yuria

Journal of Nuclear Science and Technology, 53(9), p.1342 - 1352, 2016/09

AA2015-0497.pdf:1.05MB

 Times Cited Count:3 Percentile:54.39(Nuclear Science & Technology)

Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences (AOOs) for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. The present study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was firstly defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis.

Journal Articles

Measurement of void fraction distribution in steam-water two-phase flow in a 4$$times$$4 bundle at 2 MPa

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki

Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 $$times$$ 4 bundle under 1.6 MPa (202 $$^{circ}$$C), 2.1 MPa (215 $$^{circ}$$C) and 2.6 MPa (226 $$^{circ}$$C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.

JAEA Reports

Development of BDI behavior evaluation method in the fast reactor fuel assembly; Improvement of out-of-pile bundle compression test technology

Higashiuchi, Atsushi; Ishimi, Akihiro; Katsuyama, Kozo; Uwaba, Tomoyuki; Ichikawa, Shoichi

JAEA-Technology 2015-057, 72 Pages, 2016/03

JAEA-Technology-2015-057.pdf:36.91MB

Bundle-duct interaction (BDI) in fast reactors (FRs) is one of the limiting factors for burnup. To study the high performance fuel for FR fuel, it is important to establish the method to predict accurately the BDI behavior for the fuel assembly of large-diameter fuel pins. Therefore, it was adopted a new method that the bundle compression test apparatus is placed outside the cell, the bundle specimen is put in the airtight container for contamination prevention, and the bundle specimen is carried in the cell for internal observation by X-ray CT examination apparatus. From the result of this test, it was confirmed that the new method of out-of-pile bundle compression test is carried out as it was before. The results of this test are available to study integrity assessment of fast reactor fuel, validation of the BDI analysis code and substantiation of the safety design guidelines of fast reactor. In addition, it is possible to reflect in the BDI behavior evaluation for "ASTRID".

Journal Articles

Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04

The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

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

Critical power prediction for tight lattice rod bundles

Liu, W.; Onuki, Akira; Tamai, Hidesada; Akimoto, Hajime

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

In this research, the newest version of critical power correlation for tight-lattice rod bundles is proposed by using 7-rod and 37-rod bundle data derived in Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux - critical quality type. For low mass velocity region, it is written in critical quality - annular flow length type. The correlation is verified by JAERI data and Bettis Atomic Power Laboratory data. It is confirmed the correlation is able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The correlation is further implemented into TRAC code to analyze flow decrease and power increase transients. It is confirmed transient BT can be predicted within the accuracy of the implemented critical power correlation.

Journal Articles

Proving test and analyze for critical power performance in the RMWR tight lattice rod bundles under transient condition

Liu, W.; Tamai, Hidesada; Onuki, Akira; Kureta, Masatoshi*; Sato, Takashi; Akimoto, Hajime

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 10 Pages, 2005/05

A major concern in the design of RMWR is that sufficient cooling capability be provided to keep fuel cladding temperature below specified values, even for a postulated abnormal transient process. In this research, centered the postulated transient cases that may be possibly met in the RMWR running, transient BT tests are performed in 7-rod and 37-rod double-humped tight lattice bundles, under the RMWR nominal operating condition (P = 7.2 MPa, Tin =556 K) for mass velocity G = 300 - 800 kg / (m$$^{2}$$s). Experiments are analyzed with TRAC code, in which JAERI critical power correlation is implemented for BT judgment. The traditional quasi-steady-state prediction of BT in transient process is confirmed being applicable for the postulated nominal transients in the RMWR cores.

Journal Articles

Critical power correlation for tight-lattice rod bundles

Liu, W.; Kureta, Masatoshi; Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 42(1), p.40 - 49, 2005/01

 Times Cited Count:7 Percentile:51.07(Nuclear Science & Technology)

In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For low mass velocity region ($$<$$ 300 kg/m$$^{2}$$s), the correlation is written in critical quality - annular flow length type. For high mass velocity region ($$>$$ 300 kg/m$$^{2}$$s), it is written in local critical heat flux - critical quality type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/m$$^{2}$$s and pressure from 2 to 11 MPa.

Journal Articles

Predicted two-phase flow structure in a fuel bundle of an advanced light-water reactor

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10

no abstracts in English

Journal Articles

Laser pattern profile emitted through optical glass fiber bundle with load

Yamauchi, Toshihiko

Japanese Journal of Applied Physics, 43(9A), p.6473 - 6474, 2004/09

 Times Cited Count:0 Percentile:100(Physics, Applied)

The output pattern of He-Ne laser that passes through the optical glass fiber bundle used for free electron laser etc. depends on the weight of load (modulation of transverse mode). The increment of pattern radius exponentially increases with the load. It is considered that the micro bending causes this increment. Next, the estimated delay time also exponentially increases with the weight of load, and the estimated delay time is the extent of pico-second.

Journal Articles

Critical power in 7-rod tight lattice bundle

Liu, W.; Kureta, Masatoshi; Akimoto, Hajime

JSME International Journal, Series B, 47(2), p.299 - 305, 2004/05

Experimental research on critical power in tight lattice bundle that simulates the Reduced-Moderation Water Reactor (RMWR) has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3 mm equilateral triangular pitch. Each rod is 13 mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under ${it G}$ = 100-1400 kg/m$$^{2}$$s, ${it P}$$$_{ex}$$ = 2-8.5 MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.

Journal Articles

Experimental study on thermal-hydraulics and neutronics coupling effect on flow instability in a heated channel with THYNC facility

Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), 16 Pages, 2003/10

Thermal-hydraulic and neutronic dynamics are always interrelated in BWR core. This is called thermal-hydraulic and neutronic (T/N) coupling. Channel stability experiments with T/N coupling under non-nuclear condition are very limited. This is mainly due to the difficulties in the real-time simulation of neutron dynamics and in the fast-response void fraction measurement under high-pressure and temperature conditions. Authors have developed techniques to solve the above difficulties, and have succeeded in experimentally simulating T/N coupling under non-nuclear conditions with the THYNC facility. Using THYNC facility, T/N coupling effect on channel stability was investigated. Experiments were performed under Pressure=2-7MPa, Subcooling=10-40K, and Mass flux=270-660kg/m$$^{2}$$s. THYNC results indicated T/N coupling lowered the channel stability threshold. The reduction of channel stability threshold due to T/N coupling was small within 10% at 7MPa in the present THYNC experiment, although the experimental condition was set to be more severe than that supposed in a reactor.

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

Experimental result of BWR post-CHF tests; Critical heat flux and post-CHF heat transfer coefficient (Contract research)

Iguchi, Tadashi; Iwaki, Chikako*; Anoda, Yoshinari

JAERI-Research 2001-060, 91 Pages, 2002/02

JAERI-Research-2001-060.pdf:6.34MB

no abstracts in English

Journal Articles

Analysis of multi-dimensional boiling flow in secondary water pool of horizontal PCCS; Effect of pool size

Onuki, Akira; Nakamura, Hideo; Kawamura, Shinichi*; Saishu, Sadanori*

Nippon Kikai Gakkai Netsu Kogaku Koenkai Koen Rombunshu, p.31 - 32, 2001/11

A passive containment cooling system (PCCS) is under planning to use in a next-generation-type BWR for long-term cooling by condensing steam using horizontal heat exchangers. Heat transfer behavior in a secondary water pool is one of important phenomena governing heat removal performance of the PCCS. Boiling and condensation can be supposed under high heat flux regions and the characteristics of the two-phase natural circulation should be evaluated. This study investigated effects of pool size on the characteristics by multi-dimensional two-fluid model code ACE-3D. It was found from the analyses that the pool size gives no significant influences for the characteristics in tube bundle under local-boiling mode.

JAEA Reports

Development of quick-response area-averaged void fraction meter; Application to BWR condition

Iguchi, Tadashi; Watanabe, Hironori; Kimura, Mamoru*; Anoda, Yoshinari

JAERI-Research 2001-032, 111 Pages, 2001/05

JAERI-Research-2001-032.pdf:4.14MB

no abstracts in English

52 (Records 1-20 displayed on this page)