Search Results 1-20 of 543

  • A conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

    KUBO Shigenobu , WATANABE Osamu , HIGURASHI Koichi , CHIKAZAWA Yoshitaka , OHSHIMA Hiroyuki , UCHITA Masato , MIYAGAWA Takayuki , ETO Masao , SUZUNO Tetsuji , MATOBA Ichiyo , ENDO Junji

    … <p>The authors are developing the design concept of the pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDG) for Generation IV SFRs. … The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. …

    Mechanical Engineering Journal 7(3), 19-00489-19-00489, 2020

    J-STAGE 

  • Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool (confirmation of fuel temperature calculation function with oxidation reaction in the SAMPSON code)

    SUZUKI Hiroaki , MORITA Yoshihiro , NAITOH Masanori , NEMOTO Yoshiyuki , KAJI Yoshiyuki

    … Not only the SFP but also upper spaces of the SFP, walls of the reactor building, and the blowout panel were included. … Experiments done by Sandia National Laboratory using simulated fuel assemblies equivalent to those of an actual BWR plant were analyzed by the modified SAMPSON code to confirm the function of fuel temperature calculation in the event of loss of fuel cooling in the SFP. …

    Mechanical Engineering Journal 7(3), 19-00450-19-00450, 2020

    J-STAGE 

  • Japan's Efforts to Develop the Concept of JA DEMO During the Past Decade

    飛田 健次 , 日渡 良爾 , 坂本 宜照 , 染谷 洋二 , 朝倉 伸幸 , 宇藤 裕康 , 三善 悠矢 , 徳永 晋介 , 本間 裕貴 , 角舘 聡 , 中島 徳嘉 , 原型炉設計合同特別チーム

    … This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. … Japan is currently working on the conceptual study of a steady state DEMO (JA DEMO) with a major radius of 8.5 m and fusion power of 1.5-2 GW based on water-cooled solid breeding (WCSB) blanket with PWR water condition (290-325ºC, 15.5 MPa). …

    Fusion Science and Technology (75), 372-383, 2019-06

    IR 

  • NUMERICAL ANALYSIS OF STEAM CONDENSATION AT SUB-ATMOSPHERIC PRESSURE IN WATER SUPPRESSION TANK

    Pesetti Alessio , Martelli Daniele , Frano Rosa Lo , Sarkar Biswanath , Olcese Marco , Aquaro Donato

    … A preliminary analysis has been performed with ANSYS© Fluent code to support the experimental campaign planned to be carried out at DICI-University of Pisa in the Experimental Test Tank (ETT), aiming at studying and characterizing Direct Contact Condensation (DCC) in relevant configurations (subatmospheric pressure) for International Thermonuclear Experimental Reactor (ITER) Vacuum Vessel Pressure Suppression System (VVPSS). …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2210, 2019

    J-STAGE 

  • STEAM CONDENSATION HEAT TRANSFER INSIDE REACTOR CONTAINMENT DURING THE INITIAL TRANSIENT OF A SEVERE ACCIDENT

    Punetha Maneesh , Yadav Mahesh Kumar , Bhanawat Abhinav , Khandekar Sameer

    … During the Loss of Coolant Accident (LOCA) situations, a significant amount of steam comes inside the containment. … Heat transfer is highly transient during the initiation of the accident (initial blow-down period) and occurs in the presence of high content of Non-Condensable Gases (NCGs), primarily air. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2166, 2019

    J-STAGE 

  • EFFECT OF SURFACE INCLINATION ON FILM CONDENSATION HEAT TRANSFER IN THE PRESENCE OF AIR

    Bhanawat Abhinav , Yadav Mahesh Kumar , Punetha Maneesh , Khandekar Sameer

    … Condensation of steam inside a reactor containment during Loss of Coolant Accident (LOCA) situations is essential to maintain the integrity of the containment. … In such situations, steam condenses, in the presence of Non-Condensable Gases (NCGs), on the containment walls as well as on the surfaces of different safety devices/components employed inside the containment. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2133, 2019

    J-STAGE 

  • EFFECT OF EXPERIMENTAL SETTING AND SURFACE ROUGHNESS ON OXIDATION BEHAVIOR OF ZRY-4 IN STEAM AT 1273 K

    Negyesi Martin , Amaya Masaki

    … This work deals with oxidation behavior of Zry-4 fuel cladding exposed to steam at 1273 K. … The condition corresponds to loss of coolant accident (LOCA) of the reactor core with possibility of the breakaway oxidation. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 2112, 2019

    J-STAGE 

  • AN EXPERIMENTAL RESEARCH OF THE RESISTANCE CHARACTERISTICS OF THE ANNULAR FUEL ASSEMBLY

    Duan Minghui , Zhao Minfu , Gu Hanyang

    … An annular fuel assembly includes both inner and outer coolant flow channels, so its resistance characteristics are more complex than a solid fuel assembly. … The resistance characteristics of the annular fuel assembly concern the flow distribution in the inner and outer flow channel of the assembly and the core pressure loss. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1866, 2019

    J-STAGE 

  • FLOW STUDY OF AUXILIARY FEEDWATER SYSTEM IN A NUCLEAR POWER PLANT

    Yang Yi , Yu Pei , Zhao Bin , Sun Qi , Hou Ting

    … In the event of an accident in any normal feedwater systems (Motor driven feedwater pump system, Main feedwater flow control system, Start-up feedwater system), the auxiliary feedwater system is operated, to ensure that the steam generator is supplied with an appropriate amount of water to discharge the decay heat of the core. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1767, 2019

    J-STAGE 

  • INVESTIGATIONS ON THE THERMAL-HYDRAULIC BEHAVIOR IN A GENERIC PWR KONVOI DURING A 1% COLD LEG SMALL-BREAK LOSS OF COOLANT ACCIDENT USING THE SYSTEM CODE ATHLET

    Pescador E. Diaz , Schäfer F. , Wilhelm P. , Kliem S.

    … In the presented paper, a simulation of a small-break loss of coolant accident (SBLOCA) with a 1% break in the cold leg 1 in a generic German PWR KONVOI model is carried out and investigated by means of the thermal-hydraulic system code ATHLET 3.1A. … The accident analysis is focused first on a thermal-hydraulic characterization of the SBLOCA, and a subsequent qualitative comparison with the test PKL H1.1. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1777, 2019

    J-STAGE 

  • MODIFICATION OF SEVERE ACCIDENT ANALYSIS CODE FOR ACCIDENT TOLERANT FUEL (ATF) APPLICATIONS

    Tao XU , Yirui WU , Bin ZHANG , Haifu MA , Junyun YANG , Junlong WANG , Chao TIAN , Kai WANG , Yaxin DENG

    … After the Fukushima nuclear accident, we gradually realized the safety issues of the traditional UO<sub>2</sub>+Zr nuclear fuel system. … Accident tolerance fuel (ATF) is a new generation fuel concept proposed to improve the performance of fuel elements against severe accidents, and can withstand the consequences of accidents for a long time while maintaining or improving their performance under normal operating conditions. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1939, 2019

    J-STAGE 

  • THE ANALYSIS OF LOSS OF COOLANT ACCIDENT (LOCA) INITIATING EVENT FREQUENCY

    Feng Churan , Liu Bin , Dong Fangyu , An Jin , Yang Jian

    … Probabilistic safety assessment (PSA) is an important tool to evaluate the safety of the nuclear power plant (NPP). … As the start of PSA, the quality of initiating events analysis has significant influence on NPP safety assessment and risk informed PSA application. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1837, 2019

    J-STAGE 

  • DEVELOPMENT AND PRELIMINARY SAFETY ANALYSIS OF THE SMALL MODULAR REACTOR BOC-600

    QI Zhanfei , WANG Weiwei , LIU Zhan , WANG Guodong , WANG Haitao , CAO Zhen

    … The passive safety systems and inherent safety of the BOC-600 provide a major enhancement in plant safety to passively maintain core cooling and containment integrity. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1828, 2019

    J-STAGE 

  • PARAMETERIZATION OF FLOODING IN PRESSURIZED WATER REACTOR DURING LOSS OF COOLANT ACCIDENT

    Samal Kumar , Ghosh Suman

    … Extensive numerical attempts are made here to analyze, visualize and parameterize the onset of flooding in the hot-leg of a pressurized water reactor (PWR) during loss of coolant accident. … During loss of coolant accident, counter current twophase flow may appear in the hot-leg of PWR due to the leakage in primary circuit. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1595, 2019

    J-STAGE 

  • STUDY ON LOSS-OF-COOLING AND LOSS-OF-COOLANT ACCIDENTS IN SPENT FUEL POOL (5) INVESTIGATION OF COOLING EFFECTS OF SFP SPRAY AND ALTERNATE WATER INJECTION USING MAAP CODE

    Nishimura Satoshi , Satake Masaaki , Nishi Yoshihisa , Nemoto Yoshiyuki , Kaji Yoshiyuki

    … In the present study, analyses of accident progression in a spent fuel pool (SFP) were performed using the Modular Accident Analysis Program (MAAP) code to investigate the cooling characteristics of an SFP spray and alternate water injection in loss-of-cooling and loss-of-coolant accidents. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1603, 2019

    J-STAGE 

  • RISK ANALYSIS ON CONTAINMENT BYPASS ACCIDENTS IN M310 NUCLEAR POWER PLANT

    Jian Yang , Jin Yan Du , Chu Ran Feng

    … In this paper, Level 1 and 2 PSA model of Fuqing 1 unit was built to quantify the risk of containment bypass accidents. … Containment bypass accidents are the main contributor for the LERF of M310 Nuclear Power Plant, and SGTR initiating event accounts for 76 percent in the LERF. … Finally, some targeted design improvements were discussed from the point of view of reducing risk in this paper. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1604, 2019

    J-STAGE 

  • A CONCEPTUAL DESIGN STUDY OF POOL-TYPE SODIUM-COOLED FAST REACTOR WITH ENHANCED ANTI-SEISMIC CAPABILITY

    Kubo Shigenobu , Watanabe Osamu , Higurashi Koichi , Chikazawa Yoshitaka , Ohshima Hiroyuki , Uchita Masato , Miyagawa Takayuki , Eto Masao , Suzuno Tetsuji , Matoba Ichiyo , Endo Junji

    … The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan’s specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. … The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1651, 2019

    J-STAGE 

  • DEVELOPMENT OF THERMAL-HYDRAULICS MODEL FOR TWO-PHASE FLOW IN THE HORIZONTAL RECTANGULAR FINNED CHANNEL

    Takeyama Daiki , Iwaki Chikako , Tahara Mika , Onitsuka Yoichi

    … For mitigation of severe accident with core melting, several types of core-catcher have been developed to catch and cool molten core. … It also has a flow path consisting of risers, coolant channels and downcomers to generate natural circulation flow of cooling water. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1650, 2019

    J-STAGE 

  • INVESTIGATION OF THE FLOW NEAR THE WALL OF A SINGLE IMPINGING JET AT THE SCALED UPPER PLENUM OF HTGR USING TR-PIV

    Alwafi Anas , Nguyen Thien , Hassan Yassin , Anand N.K.

    … One of the competitive candidate of the future nuclear reactors is the High Temperature Gas-cooled Reactor (HTGR). … scaled upper plenum of a HTGR was designed at Texas A&M University to study the flow behaviors in the upper plenum of the HTGR during a Loss-Of-Forced-Coolant Accident (LOFC). …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1616, 2019

    J-STAGE 

  • NUMERICAL SIMULATION OF DIRECT CONTACT CONDENSATION IN CONTAINMENT SUPPRESSION POOL WITH MELCOR

    Liu Xinxing , Sun Zhongning , Wang Jianjun , Meng Zhaoming

    … The pressure suppression pool plays an important role in BWR reactor, which was designed to have the capability as a heat sink and condenser in concern of the LOCA (loss of coolant accident). … The DCC (direct contact condensation) is the main process occurred in the suppression pool when large amount of steam was vent into the pool through the blow down pipes from the dry well. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2019.27(0), 1524, 2019

    J-STAGE 

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