| 1. |
大貫 晃, 高瀬 和之, 呉田 昌俊, 吉田 啓之, 玉井 秀定, Liu W., 秋本 肇, 高稠密格子水冷却炉心の除熱技術の開発,1; 全体計画, Proceedings of 2004 International Congress on Advances in Nuclear Power Plants (ICAPP '04), 10.1299/jsmemecjo.2003.3.0_247, 2003, 1488-1494, 2004.06, 日本原子力研究所では高稠密格子水冷却炉心(RMWR)の熱流動特性を予測する技術開発プロジェクトを電力,メーカ,大学の協力を得て平成14年度より開始した。RMWRは成熟した軽水炉技術を活用し、ウラン資源の有効利用,プルトニウムの多重リサイクル,高燃焼度,長期サイクル運転といった長期的なエネルギー供給を担える革新的な水冷却炉としての特徴を有している。RMWRは核分裂性プルトニウムの増殖比を高めるため、燃料集合体を稠密にし、ボイド率を高くしている。そのため、熱流動に関する成立性が大きな開発課題となっている。本シリーズ報告ではこの成立性にかかわる研究に焦点を当て、大型試験装置と先進的な数値解析技術を活用した研究・開発計画を述べる。. |
| 2. |
大貫 晃, 呉田 昌俊, 吉田 啓之, 玉井 秀定, Liu W., 三澤 丈治, 高瀬 和之, 秋本 肇, 稠密格子炉心熱流動特性技術開発,1; 全体計画とこれまでの成果, Journal of Power and Energy Systems (Internet), 10.1299/jpes.2.229, 2, 1, 229-239, 2008.01, R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.. |
| 3. |
MD. Iqbal Hosan, Mizuki Koga, Akihiro Kakoi, Koji Morita, Wei Liu, Xu Cheng, Experimental Study on Accident Source Terms Transport and Deposition Behavior in Nuclear Power Plants, Proceedings of the 30th International Conference on Nuclear Engineering ICONE30 May21-26, 2023, ICONE30- 1806, 2023.05. |
| 4. |
Zeren Zou, Koji Morita, Wei Liu, Development of a simplified one-dimensional CDA bubble model for source term evaluation, Proceedings of the 30th International Conference on Nuclear Engineering ICONE30 May21-26, 2023, ICONE30- 1048, 2023.05. |
| 5. |
Ting Zhang, Yao Yao, Koji Morita, Xiaoxing Liu, Wei Liu, Yuya Imaizumi, Kenji Kamiyama, A large-scale particle-based simulation of heat and mass transfer behavior in EAGLE ID1 in-pile test, Proceedings of the 30th International Conference on Nuclear Engineering ICONE30 May21-26, 2023, ICONE30- 1062, 2023.05. |
| 6. |
Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima, Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: (1) Safety Analysis of Device-Loaded Cores with Different Fuel Materials, Proceedings of the 30th International Conference on Nuclear Engineering ICONE30 May21-26, 2023, ICONE30- 1582, 2023.05. |
| 7. |
Hiroshi Sagara, Masatoshi Kawashima, Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: (2) A Study on Selecting Candidate Fuel Materials for the Basic Device Specifications, Proceedings of the 30th International Conference on Nuclear Engineering ICONE30 May21-26, 2023, ICONE30- 1811, 2023.05. |
| 8. |
Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima, Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Preliminary Study, Journal of Nuclear Engineering and Radiation Science, 10.1115/1.4056834, 9, 2, 2023.03, Abstract
Following the Fukushima Nuclear Power Plant accident in 2011, it has become increasingly important for reactor safety designs to consider measures that can prevent the occurrence of severe accidents. This report proposes a novel subassembly-type passive reactor shutdown device that expands the diversity and robustness of core disruptive accident (CDA) prevention strategies for sodium-cooled fast reactors. The developed device contains pins with a fuel material that is in the solid state during normal operation but melts into a liquid when the temperature exceeds a certain value (i.e., during a potential accident). When an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident occurs, the device can passively provide significant negative reactivity by rapidly transferring liquefied device fuel into the lower plenum region of the pins via gravitation alone. The reactors containing some of the proposed devices in place of original fuel subassemblies become subcritical before the driver fuels are damaged, even if ULOF or UTOP transient events occur. The present study evaluates candidate materials for device fuels (e.g., metallic alloy, chloride), optimal device pin structures for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions to demonstrate the engineering applicability of the proposed device. This report discusses preliminary results regarding the nuclear requirements for inducing negative reactivity to achieve reactor shutdown under the expected device conditions during an accident.. |
| 9. |
Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima, Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications, Journal of Nuclear Engineering and Radiation Science, 10.1115/1.4056854, 9, 4, 2023.03, Abstract
A new subassembly type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device passively provides high negative reactivity to the core. This study evaluated the nuclear and thermal properties of the device subassembly with metallic fuel to determine the device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWel-class mixed-oxide-fueled SFR core confirmed that a conventional homogeneous core maintains stable cooling of the core before coolant boiling in the driver fuel subassemblies. In contrast, the negative reactivity required to terminate the event by device operation was slightly higher in the low sodium void reactivity core than in the conventional homogeneous core.. |
| 10. |
Ting Zhang, Koji Morita, Xiaoxing Liu, Wei Liu, Kenji Kamiyama, A 3D particle-based simulation of heat and mass transfer behavior in the EAGLE ID1 in-pile test, Annals of Nuclear Energy, 10.1016/j.anucene.2022.109389, 179, 15, 109389-109389, 2022.12. |
| 11. |
M. Koga, K. Morita, W. Liu, T. Matsumoto, K. Takanishi, K. Nakamura, T. Kanai, Experimental Study on Aerosol Migration Behavior in Rectangular Penetrations, 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety, N12P1042, 2022.10. |
| 12. |
W. Liu, On Mechanistic Prediction of Critical Heat Flux for Nuclear Power Plants (5) Mechanistic Models for Critical Heat Flux Prediction in Subcooled Flow boiling, 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety, N12E1048, 2022.10. |
| 13. |
D. Takatsuka, K. Morita, T. Nakamura, T. Zhang, W. Liu, K. Kamiyama, Particle-based Simulation of Jet Impingement Behaviors, 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety, N12P1046, 2022.10. |
| 14. |
Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima, DEVELOPMENT OF A PASSIVE REACTOR SHUTDOWN DEVICE TO PREVENT CORE DISRUPTIVE ACCIDENTS IN FAST REACTORS: A STUDY ON BASIC DEVICE SPECIFICATIONS, Proceedings of the 2022 29th International Conference on Nuclear Engineering ICONE29 August 8-12, 2022, ICONE29-91812, 2022.08. |
| 15. |
Wei Liu, Prediction of Critical Heat Flux for Subcooled Flow Boiling in Annulus and Transient Surface Temperature Change at CHF, Fluids, 10.3390/fluids7070230, 7, 7, 230-230, 2022.07, The ability to predict critical heat flux (CHF) is of considerable interest for high-heat equipment, including nuclear reactors. CHF prediction from a mechanistic model for subcooled flow boiling in rod bundles still remains unsolved. In this paper, we try to predict the CHF in an annulus, which is the most basic flow geometry simplified from a fuel bundle, using a liquid sublayer dryout model. The prediction is validated with both water and R113 data, showing an accuracy within ±30%. After the CHF in an annulus is calculated successfully, a near-wall vapor–liquid structure is proposed on the basis of the liquid sublayer dryout model. Modeling of heat transfer modes over the heating surface at CHF is performed, and predictions of the changes in liquid sublayer thickness and heater surface temperature at the CHF occurrence point are carried out by solving the heat conduction equation in cylindrical coordinates with a convective boundary condition, which changes with the change in flow pattern over the heating surface. Transient changes in the liquid sublayer thickness and surface temperature at the CHF occurrence point are reported.. |
| 16. |
Wei Liu, Kazuya Gotou, Akihiro Endo, Tsutaya Matumoto and Koji Morita, FLOW CHARACTERISTICS IN RECTANGULAR MICRO-CHANNELS WITH HIGH ASPECT RATIOS, The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), 35887, 2022.03. |
| 17. |
Ting ZHANG, Koji MORITA, Wei LIU, Xiaoxing LIU, Kenji KAMIYAMA, Numerical investigation on mechanism of heat transfer between molten pool and duct wall in EAGLE ID1 and ID2 in-pile tests, The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), 33908, 2022.03. |
| 18. |
Md Abdur Rafiq Akand, Tatsuya Matsumoto, Wei Liu, Koji Morita, Mechanistic critical heat flux prediction for in-vessel retention conditions, Nuclear Engineering and Design, 10.1016/j.nucengdes.2021.111494, 384, 111494-111494, 2021.12. |
| 19. |
日本原子力学会, 原子炉における機構論的限界熱流束評価技術, 研究専門委員会, 原子炉における機構論的限界熱流束評価技術の確立に向けて Part2: 機構論的限界熱流束予測評価手法確立に向けた研究とその課題、Ⅱ．これまでの限界熱流束のメカニズムと評価手法に関する研究, 10.3327/jaesjb.63.12_820, 63, 12, 820-824, 2021.12. |
| 20. |
Kuniyoshi Takamatsu, Tatsuya Matsumoto, Wei Liu, Koji Morita, Comparisons between passive RCCSs on degree of passive safety features against accidental conditions and methodology to determine structural thickness of scaled-down heat removal test facilities, Annals of Nuclear Energy, 10.1016/j.anucene.2021.108512, 162, 108512-108512, 2021.11. |
| 21. |
M.A. Rafiq Akand, Kei Kitahara, Tatsuya Matsumoto, Wei Liu, Koji Morita, Experimental study and modeling of bubble lift-off diameter in subcooled flow boiling including the inclination effect of the heating surface, Journal of Nuclear Science and Technology, 10.1080/00223131.2021.1931518, 58, 11, 1195-1209, 2021.11. |
| 22. |
Kazuya GOTO, Wei LIU, Tatsuya MATSUMOTO, Koji MORITA, FLOW CHARACTERISTICS IN RECTANULAR MICROCHANNELS, the Second Asian Conference on Thermal Sciences, ACTS-1223, 2021.10. |
| 23. |
Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Koharu Kawase, Isamu Sato, Haruaki Matsuura T, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima, Development of a Passive Reactor Shutdown Device for Prevention of Core Disruptive Accidents in Fast Reactors: Project Overview and Preliminary Results, 2021.08. |
| 24. |
Ting Zhang, Koji Morita, Xiaoxing Liu, Wei Liu, Kenji Kamiyama, A 3d Numerical Simulation on Heat Transfer Behavior in Eagle Id1 In-Pile Test Using Finite Volume Particle Method, ICONE28- 61469, 2021.08. |
| 25. |
Md. Abdur, Rafiq Akand, Kei Kitahara, Tatsuya Matsumoto, Wei Liu, Koji Morita, A Modify Model for the Net Vapor Generation Point and Its Application on CHF Prediction in Subcooled Flow Boiling, ICONE28-64022, 2021.08. |
| 26. |
Kuniyoshi Takamatsu, Tatsuya Matsumoto, Wei Liu, Koji Morita, Comparison Between Passive Reactor Cavity Cooling Systems Based on Atmospheric Radiation and Atmospheric Natural Circulation, Annals of Nuclear Energy, 10.1016/j.anucene.2020.107867, 151, p.107867_1 --p.107867_11, 2021.02. |
| 27. |
T. Zhang, K. Funakoshi, X. Liu, W. Liu, K. Morita, K. Kamiyama, Numerical Simulation of Heat Transfer Behavior in EAGLE ID1 In-Pile Test Using Finite Volume Particle Method, Ann. Nucl. Energy, 10.1016/j.anucene.2020.107856, 150, 107856-107856, 107856, 2021.01. |
| 28. |
M. A. Rafiq Akand, T. Matsumoto, W. Liu, K. MoritaM, A Modified Liquid Sublayer Dryout Model For Subcooled Flow Boiling Critical Heat Flux Prediction in IVR Condition, International Topical Meeting on Advances in Thermal Hydraulics (ATH'2020 topical meeting), ATH'2020 topical meeting Proceedings 32842, 1074-1087, 32842, 2020.10. |
| 29. |
Kuniyoshi Takamatsu, Tatsuya Matsumoto, Wei Liu, Koji Morita, Comparative Methodology between Actual RCCS and Downscaled Heat-Removal Test Facility, Annals of Nuclear Energy, 133, 11, 830-836, vol. 133, pp 830-836 (November 2019)., 2019.11. |
| 30. |
Le Hoang Sang Phan, Phi Manh Ngo, Ryo Miura, Yusuke Tasaki, Tatsuya Matsumoto, Wei Liu & Koji Morita, Self-leveling behavior of mixed solid particles in a cylindrical bed using a gas-injection method, Journal of Nuclear Science and Technology, 56, 1, 111-122, 2019.01. |
| 31. |
Kuniyoshi Takamatsu, Tatsuya Matsumoto, Wei Liu, Koji Morita, Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection, Annals of Nuclear Energy, 10.1016/j.anucene.2018.08.047, 122, 201-206, 2018.12. |
| 32. |
Seisuke Hosomi, Tomoyasu Akashi, Koji Morita, Wei Liu, Tsutaya Matsumoto, Kinuyoshi Takamatsu, Experimental study on heat removal performance of a new reactor cavity cooling system (RCCS), Proceedings of 11th Korea –Japan Symposium on Nuclear Thermal Hydraulics and Safety, No.0144, 2018.11. |
| 33. |
Masatsugu Kato, Kanji Funakoshi, Xiao Xing Liu, Tatsuya Matsumoto, Wei Liu and Koji Morita, Validation of a three-dimensional finite-volume-particle method for simulation of liquid-liquid mixing flow behavior, Proceedings of 11th Korea – Japan Symposium on Nuclear Thermal Hydraulics and Safety, No. 0142, 2018.11. |
| 34. |
Wei Liu, Predictions of Critical Heat Flux for Subcooled Flow Boiling in Annulus, Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12), No.1148, 2018.10. |
| 35. |
Le Hoang, Sang PHAN, Yohei OHARA, Ry KAWATA, Xiaoxing LIU, Wei LIU, Koji MORITA, Liancheng GUO, Kenji KAMIYAMA, Hirotaka TAGAMI, Numerical Simulation on Self-leveling Behavior of Mixed Particle Beds Using Multi-fluid Model Coupled with DEM, Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12), No.1065, 2018.10. |
| 36. |
Koji Morita, Ryusei Ogawa, Hiromi Tokioka, Xaoxing Liu, Wei Liu, Kenji Kamiyama, Particle-based Simulation of Heat Transfer Behavior in EAGLE ID1 In-pile Test, Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12), No.954, 2018.10. |
| 37. |
Wei Liu, Prediction of Transient Surface Temperature Changes at Subcooled Flow Boiling DNB, the 10th International Conference on Boiling and Condensation Heat Transfer, A360, 2018.03. |
| 38. |
上澤 伸一郎, 劉 維, 焦 利芳, 永武 拓, 高瀬 和之, 柴田 光彦, 吉田 啓之, 内管加熱二重管における海水の非沸騰熱伝達への影響, 日本原子力学会和文論文誌, 10.3327/taesj.j15.024, 15, 4, 183-191, 2016.04, Seawater was injected into the reactors during the accident at TEPCO's Fukushima Daiichi NPS. However, the effects of the seawater on the cooling performance of the fuel rods and fuel debris are not clear. As possible effects, the change in the physical properties of the coolant and the sea salt deposition on a heat transfer surface and in the coolant are considered. We conducted thermal-hydraulic experiments using an internally heated annulus to determine the effects of seawater under conditions without boiling. The same experiments for water and sodium chloride (NaCl) solution were also conducted for the purpose of comparison with the artificial seawater. In these experiments, considering the physical properties of the artificial seawater, the thermal-hydraulic behaviors of the artificial seawater under forced convection (Re>2300 [－]) was estimated from the Dittus-Boelter correlation although sea salt was deposited in the fluid. According to the results of particle image velocimetry (PIV), the velocity distribution in the artificial seawater was NOT different from that in the water and the NaCl solution. For a mixed convection regime, the Nusselt number of the artificial seawater was obtained from the correlation of the Grasholf number, Reynolds number and Prandtl number, as well as those for the water and the NaCl solution. Therefore, considering the physical properties of the artificial seawater, the thermal-hydraulic behavior of the seawater in single-phase flow can be estimated from the conventional thermal-hydraulic correlations for a single-phase flow. . |
| 39. |
Wei Liu, Hidesada Tamai, Kazuyuki Takase, Pressure Drop and Void Fraction in Steam-Water Two-Phase Flow at High Pressure, Journal of Heat Transfer, 10.1115/1.4023678, 135, 8, 2013.08, For a steam generator (SG) in a commercialized sodium-cooled fast breeder reactor (FBR), flow instability in the water side is one of the most important items needing research. As the first step of this research, thermal-hydraulic experiments using water as the test fluid were performed under high pressure conditions at the Japan Atomic Energy Agency (JAEA) by using a circular tube. Void fraction, pressure drop, and heat transfer coefficient data were obtained under 15, 17, and 18 MPa. This paper discusses the steam-water pressure drop and void fraction. Using the obtained data, we evaluated existing correlations for void fraction and two-phase flow multipliers under high pressure. As a result, the drift flux model implemented in the TRAC-BF1 code was confirmed to suitably predict the void fraction well under the present high pressure conditions. For the two-phase flow multiplier, the Chisholm correlation and the homogeneous model were confirmed to be the best under the present high-pressure conditions.. |
| 40. |
Wei Liu, Kazuyuki Takase, Development of measurement technology for surface heat fluxes and temperatures, Nuclear Engineering and Design, 10.1016/j.nucengdes.2011.06.036, 249, 166-171, 2012.08. |
| 41. |
LIU Wei, HIDESADA Tamai, TAKASE Kazuyuki, HAYAFUNE Hiroki, FUTAGAMI Satoshi, KISOHARA Naoyuki, Steam Water Pressure Drop under 15 MPa, Journal of Power and Energy Systems, 10.1299/jpes.5.229, 5, 3, 229-240, 2011.03, For a steam generator with straight double-walled heat transfer tubes that will be used in a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15MPa. Several two-phase flow multipliers were checked and then, it was found that both Chisholm two-phase flow multiplier and homogeneous model can predict the present experimental data in high accuracy. A sudden decrease of the pressure drop was observed when flow pattern shifts from bubbly and churn flows to annular flow. The reason for this decrease is tried to be interpreted.. |
| 42. |
LIU Wei, KURETA Masatoshi, TAKASE Kazuyuki, Experimental Research on the Effect of Axial Power Distribution on Critical Power, Journal of Power and Energy Systems, 10.1299/jpes.3.301, 3, 1, 301-312, 2009.01, This paper concerns experimental research to ascertain the effect of axial power distribution on critical power in the positive quality region. Experiments took place at atmospheric pressure in a circular tube. Axial uniform heating and two other axial non - uniform heating cases were selected for detailed evaluation. The effects of relative power ratio on critical power, critical quality and critical boiling length were ascertained in detailed evaluations. Using the experimental data, we evaluated existing correlating concepts with critical power. Result showed a combination of the overall power concept (χBT - LB) and the local conditions concept (χBT - qBT) appearing to be promising in correlating present critical power data in axial non - uniform heating conditions.. |
| 43. |
Wei Liu, Akira Ohnuki, Hiroyuki Yoshida, Masatoshi Kureta, Kazuyuki Takase, Hajime Akimoto, Thermal Feasibility Analyses for the 1356MWe High Conversion-Type Innovative Water Reactor for Flexible Fuel Cycle, Heat Transfer Engineering, 10.1080/01457630801981614, 29, 8, 704-711, 2008.08. |
| 44. |
Hidesada TAMAI, Masatoshi KURETA, Wei LIU, Takashi SATO, Toru NAKATSUKA, Akira OHNUKI, Hajime AKIMOTO, Effect of Rod Bowing on Critical Power;Based on Tight-Lattice; Rod Bundle Experiments, Journal of Nuclear Science and Technology, 45, 6, 567-574, 2008.02. |
| 45. |
LIU Wei, TAMAI Hidesada, KURETA Masatoshi, OHNUKI Akira, AKIMOTO Hajime, Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions, Journal of Power and Energy Systems, 10.1299/jpes.2.240, 2, 1, 240-249, 2008.01, A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.. |
| 46. |
Liu Wei, Kureta Masatoshi, Tamai Hidesada, OHNUKI Akira, AKIMOTO Hajime, Critical Power Characteristics in 37-rod Tight Lattice Bundles under Transient Conditions, Journal of nuclear science and technology, 10.1080/18811248.2007.9711360, 44, 9, 1172-1181, 2007.09. |
| 47. |
Liu Wei, Kureta Masatoshi, Yoshida Hiroyuki, OHNUKI Akira, AKIMOTO Hajime, An Improved Critical Power Correlation for Tight-Lattice Rod Bundles, Journal of nuclear science and technology, 10.1080/18811248.2007.9711845, 44, 4, 558-571, 2007.04. |
| 48. |
Tamai Hidesada, Kureta Masatoshi, Liu Wei, SATO Takashi, OHNUKI Akira, AKIMOTO Hajime, Gap Width Effect on Critical Power based on Tight-Lattice 37-Rod Bundle Experiments, Journal of nuclear science and technology, 10.1080/18811248.2007.9711256, 44, 1, 54-63, 2007.01. |
| 49. |
Kureta Masatoshi, Tamai Hidesada, Ohnuki Akira, SATO Takashi, LIU Wei, AKIMOTO Hajime, Critical Power Experiment with a Tight-Lattice 37-Rod Bundle, Journal of nuclear science and technology, 10.1080/18811248.2006.9711082, 43, 2, 198-205, 2006.02. |
| 50. |
M. Monde, W. Liu, Y. Mitsutake, Kyaw Zin Oo, Characteristics of boiling curve in transition region between nucleate boiling and film boiling, Heat Transfer—Asian Research, 10.1002/htj.20097, 35, 1, 20-34, 2006.01. |
| 51. |
門出 政則, 劉 維, 光武 雄一, Kyaw Zin OO, 遷移域の沸騰曲線の特性, 日本機械学会論文集B編, 10.1299/kikaib.71.1390, 71, 705, 1390-1397, 2005.05. |
| 52. |
Wei Liu, Hideki Nariai, Ultrahigh CHF Prediction for Subcooled Flow Boiling Based on Homogenous Nucleation Mechanism, Journal of Heat Transfer, 10.1115/1.1844536, 127, 2, 149-158, 2005.02, Homogeneous nucleation, although being discounted as a mechanism for vapor formation for water in most conditions, is found to possibly occur under some extreme conditions in subcooled flow boiling. Under the conditions, vapor bubbles of molecular dimensions generated in the superheated liquid adjacent to channel wall from homogeneous nucleation due to the local temperature exceeds homogeneous nucleation temperature. The condition is called in this paper as homogeneous nucleation governed condition. Under the condition, conventional flow pattern for subcooled flow boiling, which is characterized by the existence of Net Vapor Generation (NVG) point and the followed bubble detachment, movement and coalescence processes, cannot be established. Critical heat flux (CHF) triggering mechanism so far proposed, which employs a premise assumption that the conventional flow pattern has been established, such as liquid sublayer dryout model, is no more appropriate for the homogeneous nucleation governed condition. In this paper, first, the existence of the homogeneous nucleation governed condition is indicated. In the following, a criterion is developed to judge a given working condition as the conventional one or the homogeneous nucleation governed one. With the criterion, subcooled flow boiling data are categorized and typical homogeneous nucleation governed datasets are listed. The homogeneous nucleation governed data are characterized by extreme working parameters, such as ultrahigh mass flux, ultralow ratio of heated length to channel diameter L/D or ultrahigh pressure. CHF triggering mechanism for the homogeneous nucleation governed condition is proposed and verified. Parametric trends of the CHF, in terms of mass flux, pressure, inlet subcooling, channel diameter, and the ratio of heated length to diameter are also studied.. |
| 53. |
Liu Wei, Kureta Masatoshi, Ohnuki Akira, AKIMOTO Hajime, Critical Power Correlation for Tight-Lattice Rod Bundles, Journal of nuclear science and technology, 10.1080/18811248.2005.9726362, 42, 1, 40-49, 2005.01. |
| 54. |
劉 維, 呉田 昌俊, 玉井 秀定, 光武 徹, 大貫 晃, 秋本 肇, 稠密格子体系における過渡限界出力試験と解析(水冷却炉,革新型原子炉の開発および多目的利用技術,原子力要素技術開発), 年次大会講演論文集, 10.1299/jsmemecjo.2004.3.0_231, 2004, 231-232, 2004.09, 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, power increase and flow decrease transient tests are performed in 7-rod and 37-rod double-humped tight lattice bundles, under RMWR nominal operating condition (P_ = 7.2 MPa, T_ =283℃) for mass velocity G = 300, 450, 600 kg/m^2s. Experiments are analyzed with TRAC code, in which new JAERI critical power correlation is implemented for BT judgment. For the postulated nominal power increase and flow decrease transients, when CPR is 1.3, no Boiling Transitions (BTs) are observed in experiments and TRAC code predicts the same trends. For severer conditions that BT occurs, wall temperature jumping points (BT points) can be predicted quite well within the accuracy of the implemented critical power correlation. The traditional quasi-steady-state prediction of BT in transient process is confirmed being applicable for axially double-humped-heated tight lattice bundles.. |
| 55. |
Liu Wei, Kureta Masatoshi, Akimoto Hajime, Critical Power in 7-Rod Tight Lattice Bundle, JSME international journal. Ser. B, Fluids and thermal engineering, 10.1299/jsmeb.47.299, 47, 2, 299-305, 2004.02, The Reduced-Moderation Water Reactor (RMWR) has recently becomes of great concern. The RMWR is expected to promote the effective utilization of uranium recourse. The RMWR is based on water-cooled reactor technology, with achieved under lower core water volume and water flow rate. In comparison with the current light water reactors whose water-to-fuel volume ratio is about 2-3, in the RMWR, this value is reduced to less than 0.5. Thereby, there is a need to research its cooling characteristics. Experimental research on critical power in tight lattice bundle that simulates the 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 14.3mm equilateral triangular pitch. Each rod is 13mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under G=100-1400kg/m2s, Pex=2-8.5MPa. 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.. |
| 56. |
Masanori Monde, Hirofumi Arima, Wei Liu, Yuhichi Mitsutake, J.A. Hammad, Analytical method of two-dimensional inverse heat conduction problem using Laplace transformation: Effect of number of measurement points, Heat Transfer?Asian Research, 10.1002/htj.10116, 32, 7, 618-629, 2003.11. |
| 57. |
Masanori Monde, Wei Liu, Hirofumi Arima, Yuhichi Mitsutake, J.A. Hammad, Improvement of inverse heat conduction solution using Laplace transformation: Method of partial division of time, Heat Transfer?Asian Research, 10.1002/htj.10117, 32, 7, 630-638, 2003.11. |
| 58. |
呉田 昌俊, 劉 維, 玉井 秀定, 大貫 晃, 秋本 肇, 2237 高稠密格子水冷却炉心の除熱技術の開発 (2) : 大型熱特性試験とモデル実験, 年次大会講演論文集, 10.1299/jsmemecjo.2003.3.0_249, 2003, 249-250, 2003.08, In this manuscript, the key point of large-scale thermal-hydraulic tests and model experiments which simulate the Reduced-Moderation Water Reactor (RMWR) core are reported. The aim of the large-scale thermal-hydraulic tests using 37-rod bundle test facility is to make clear the rod number effect, rod gap effect etc. for investigating the feasibility of RMWR. Thermal-hydraulic model experiments will be performed in order to verify an advanced 3-D two-phase flow simulation method. Void fraction distribution and its fluctuation and basic thermal characteristics in tight lattice rod bundles will be measured by high-frame-rate neutron radiography technique etc. We will develop the database by 2007 that can resolve the fundamental feasibility subjects for the RMWR.. |
| 59. |
門出政則, 劉維, 光武雄一,井孝善, 逆問題解を利用した移動熱源位置の推定, 日本機械学会論文集 B編, 10.1299/kikaib.69.1651, 69, 683, 1651-1658, 2003.07. |
| 60. |
Masanori Monde, Hirofumi Arima, Wei Liu, Yuhichi Mitutake, Jaffar A. Hammad, An analytical solution for two-dimensional inverse heat conduction problems using Laplace transform, International Journal of Heat and Mass Transfer, 10.1016/s0017-9310(02)00510-0, 46, 12, 2135-2148, 2003.06. |
| 61. |
呉田 昌俊, 劉 維, 岩村 公道, 秋本 肇, A254 稠密バンドル内限界熱流束 (1) : 出力分布や流動パラメータの影響, 熱工学講演会講演論文集, 10.1299/jsmeptec.2002.0_325, 2002, 325-326, 2002.11, CHF experiments have been conducted using the rod bundles of a narrow triangular arrangement, which simulate the Reduced-Moderation Water Reactor core. The purposes of these experiments are (a) to investigate the parameter effects on critical power, (b) to evaluate the existing correlations, (c) to propose the correlation and (e) to make use of the data for the verification of the numerical analysis code. In this paper, parameter effects on critical power or critical quality were focused as a fundamental understanding. It was found from the comparison between the axially uniform and the double humped power distribution that the critical quality of the double humped one decreases significantly when mass velocity>about 150 kg/m^2s.. |
| 62. |
劉 維, 呉田 昌俊, 岩村 公道, 秋本 肇, A255 稠密バンドル内限界熱流束 (2) : 稠密バンドル限界熱流束相関式の検証, 熱工学講演会講演論文集, 10.1299/jsmeptec.2002.0_327, 2002, 327-328, 2002.11, Reduced-Moderation Water Reactor (RMWR) will be operated under low core water volume and low water flow rate. In RMWR, Water-to-fuel volume ratio (V_m/V_f) is reduced less than 0.5,in comparison with the values of 2∿3 in the current light water reactors. Thereby, there is a need to research the cooling limit. Experimental research on CHF in tight lattice bundles that simulates the actual RMWR have been carried out in JAERI. This paper will focus on the verification of existing CHF correlation for the tight lattice bundle. The Arai correlation is verified with BAPL data and JAERI data and is found that it can not give satisfied CHF prediction to the RMWR working condition.. |
| 63. |
門出政則, 有馬博史, 劉維, 光武雄二, J. A. Hammad, ラプラス変換を用いた2次元非定常熱伝導の逆問題解析 : 測定点数と内挿方法について, 日本機械学會論文集 B編, 68, 672, 2306-2312, 2002.08. |
| 64. |
門出 政則, 劉 維, 有馬 博史, 光武 雄二, HAMMAD Jaffar A, ラプラス変換を用いた熱伝導の逆問題解の改善 : 時間区分法, 日本機械学会論文集 B編, 10.1299/kikaib.68.2093, 68, 671, 2093-2097, 2002.07. |
| 65. |
W. Liu, H. Nariai, Viewpoint of Subcooled Flow Boiling Critical Heat Flux Mechanism, Chemical Engineering & Technology, 10.1002/1521-4125(200204)25:43.0.co;2-p, 25, 4, 447-453, 2002.04. |
| 66. |
W. Liu, H. Nariai, F. Inasaka, Prediction of critical heat flux for subcooled flow boiling, International Journal of Heat and Mass Transfer, 10.1016/s0017-9310(99)00373-7, 43, 18, 3371-3390, 2000.09. |
