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JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

Journal Articles

Progress of design and related researches of sodium-cooled fast reactor in Japan

Kamide, Hideki; Sakamoto, Yoshihiko; Kubo, Shigenobu; Oki, Shigeo; Ohshima, Hiroyuki; Kamiyama, Kenji

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development related to safety enhancement and the severe accident countermeasures. For the purpose of strengthening of decay heat removal function, several researches have been carried out on the decay heat removal in a core disruptive accident (CDA), diversity and applicability of decay heat removal systems, and thermal hydraulic evaluation methods. In order to elucidate the behavior of molten fuel during CDA, some in-pile and out-of-pile tests has been performed by international collaboration including basic experiments. Core design was also improved from the viewpoint of preventing the occurrence of severe accident.

Journal Articles

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Nuclear Engineering and Design, 312, p.30 - 41, 2017/02

 Times Cited Count:1 Percentile:64.68(Nuclear Science & Technology)

In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

Journal Articles

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor; Influence of flow collector of backup CR guide tube

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.

Journal Articles

An Experimental study on natural circulation decay heat removal system for a loop type fast reactor

Ono, Ayako; Kamide, Hideki; Kobayashi, Jun; Doda, Norihiro; Watanabe, Osamu*

Journal of Nuclear Science and Technology, 53(9), p.1385 - 1396, 2016/09

 Times Cited Count:2 Percentile:57.25(Nuclear Science & Technology)

Decay heat removal by natural circulation is a significant passive safety measure of a fast reactor against station blackout. The decay heat removal system (DHRS) of the loop type sodium fast reactor being designed in Japan comprises a direct reactor auxiliary cooling system and primary reactor auxiliary cooling system (PRACS). The thermal hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant. The experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, the experiments varying the pressure loss coefficients of the loop as the experimental parameters showed robustness of the PRACS.

Journal Articles

A Study on the thermal-hydraulics in the damaged subassemblies under the operation of decay heat removal system

Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04

Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.

Journal Articles

Proposal of benchmark problem of thermal striping phenomena in planar triple parallel jets tests for fundamental code validation in sodium-cooled fast reactor development

Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08

Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.

Journal Articles

Water experiments on thermal striping in reactor vessel of Japan Sodium-cooled Fast Reactor; Countermeasures for significant temperature fluctuation generation

Kobayashi, Jun; Ezure, Toshiki; Kamide, Hideki; Oyama, Kazuhiro*; Watanabe, Osamu*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. High cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60$$^{circ}$$ sector model of the reactor upper plenum. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP.

Journal Articles

Sodium experiments on natural circulation decay heat removal and 3D simulation of plenum thermal hydraulics

Kamide, Hideki; Ono, Ayako; Kimura, Nobuyuki; Endo, Junji*; Watanabe, Osamu*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13), Companion CD (CD-ROM), 11 Pages, 2015/04

Natural circulation decay heat removal is one of the significant issues for fast reactor safety, especially in long term station blackout events. Several sodium experiments were carried out using a 7-subassmbly core model for core thermal hydraulics under natural circulation conditions and for onset transients of natural circulation in a decay heat removal system (DHRS) including natural draft. Significant heat removal via inter-wrapper flow was confirmed in the experiments. Solidification of sodium in an air cooler is one of key issues in loss of heat sink events. Natural circulation characteristics under long-term decay heat removal were also obtained. Multi-dimensional phenomena, e.g., thermal stratification and bypass flow in plenums and/or heat exchangers, may influence the natural circulation. Thus, 3D simulation method was developed for entire region in the primary loop. Comparison of temperature distributions in a DHRS heat exchanger between experiment and analysis was done.

Journal Articles

Development of transfer pot for JSFR ex-vessel fuel handling

Hirata, Shingo; Chikazawa, Yoshitaka; Kato, Atsushi; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

In Fast Reactor Cycle Technology Development (FaCT) project, Japan Sodium-cooled Fast Reactor (JSFR) is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a pot which contains two fuel subassemblies simultaneously and is applicable size to compact reactor vessel, has been developing so as to shorten a refueling period leading to an improvement of plant availability. The pot is required to provide sufficient cooling capability even in case of transportation malfunction during transportation of spent fuel subassemblies with high decay heat. In the present study, experimental and analytical studies to evaluate the cooling capacity of the pot are summarized.

Journal Articles

Sodium experiments of buoyancy-driven penetration flow into low-power subassemblies in a sodium-cooled fast reactor during natural circulation decay heat removal

Kamide, Hideki; Kobayashi, Jun; Hayashi, Kenji

Nuclear Technology, 175(3), p.628 - 640, 2011/09

 Times Cited Count:1 Percentile:86.64(Nuclear Science & Technology)

Natural circulation has a significant role in the decay heat removal function of a sodium cooled reactor. The dipped heat exchanger (DHX) immersed in a reactor upper plenum provides cold sodium in the upper plenum during the decay heat removal operation. This cold sodium covers the top of the core under the natural circulation. Sodium experiments were carried out to find onset condition and penetration depth of such partial reverse flow driven by buoyancy force. A blanket subassembly and the upper plenum were modeled in the test section including an axial upper neutron shielding of the subassembly. The experimental parameters were temperature difference between hot upward flow in the channel and cold fluid in the upper plenum and flow velocity in the channel. The onset conditions of the penetration flow were correlated with Gr and Re numbers as well as basic water experiments. The observed penetration depths were limited in the upper axial neutron shielding of the subassembly.

Journal Articles

Sodium experiment on fully natural circulation systems for decay heat removal in Japan sodium-cooled fast reactor

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Watanabe, Osamu*

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 16 Pages, 2011/09

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan sodium cooled fast reactor (JSFR). Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. Influences of the pressure loss coefficient in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS due to recovery of natural circulation head via the increase of temperature difference in each loop.

Journal Articles

Sodium experiments on decay heat removal system of Japan sodium cooled fast reactor; Start-up transient of decay heat removal system

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Watanabe, Osamu*

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The JSFR has two loops of the main heat transport system in order to reduce number of components and the construction cost. The DHRS of JSFR consists of one DRACS and two PRACS, which have a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and also start-up transient of the DHRS loop. Heat transfer coefficient on the tube outer surface was in good agreement with a conventional correlation under operation condition in the reactor. The transient experiments for the start-up of DHRS loop showed that smooth increase of natural draft in the air duct followed by the sodium flow rate in the DHRS loop. Some delay of the flow rate increase was recognized in the DHRS loop as compared with that of the natural draft in the air cooler.

Journal Articles

Experimental studies of natural circulation decay Heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

Kamide, Hideki; Miyakoshi, Hiroyuki; Watanabe, Osamu*; Eguchi, Yuzuru*; Koga, Tomonari*

Nippon Kikai Gakkai Rombunshu, B, 76(763), p.460 - 462, 2010/03

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS are performed. Water experiments are carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increase of natural circulation flow rates in all systems of air and sodium was confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan.

Journal Articles

Sodium experiments of buoyancy driven penetration flow into low power subassemblies in a sodium cooled fast reactor during natural circulation decay heat removal

Kamide, Hideki; Kobayashi, Jun; Hayashi, Kenji

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 17 Pages, 2009/09

Natural circulation has a significant role in the decay heat removal function of a sodium cooled reactor. The dipped heat exchanger (DHX) immersed in a reactor upper plenum as the decay heat removal system provides cold sodium in the upper plenum during the decay heat removal operation. This cold sodium covers the top of the core under low flow rate conditions of the natural circulation. Sodium experiments were carried out to find onset condition and penetration depth of such partial reverse flow driven by buoyancy force. The onset conditions of the penetration flow were correlated with Gr and Re numbers as well as basic water experiments. The observed penetration depths were limited in the upper axial neutron shielding of the subassembly.

Journal Articles

Experimental studies of natural circulation decay heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

Kamide, Hideki; Miyakoshi, Hiroyuki; Watanabe, Osamu*; Eguchi, Yuzuru*; Koga, Tomonari*

Dai-14-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.427 - 428, 2009/06

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS are performed. Water experiments were carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increase of natural circulation flow rates in all systems of air and sodium were confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan.

Oral presentation

Scaling method for analysis model and experiment for safety study of advanced reactors, 1; Thermal hydraulics and safety of core and components in fast reactors

Kamide, Hideki

no journal, , 

Several model experiments were carried out to see phenomena and to get validation data for numerical simulation methods as for the natural circulation decay heat removal in a sodium cooled fast reactor. Here, strategy of the model experiments is summarized. When sodium is used as a working fluid, real scaled model is needed in order to match all similarity rules. The sodium experiment is applied to heat transfer oriented phenomena. As for the water experiment, thermal property is much different from that of sodium. However, this will help to mach one of the similarity rules in a small scaled apparatus. Thus, combination of the sodium experiments and the water experiments is used according to the selected phenomena.

Oral presentation

Development of evaluation methods for decay heat removal by natural circulation under transient conditions, 11; Sodium test for decay heat removal by natural circulation, 1

Kamide, Hideki; Kimura, Nobuyuki; Hayashi, Kenji; Miyakoshi, Hiroyuki; Watanabe, Osamu*

no journal, , 

Sodium experiments were carried out for the natural circulation decay heat removal in a large scale sodium cooled fast reactor (JSFR) developed by FaCT project. The purposes of the sodium experiments are to grasp the transient characteristics of soium flow rate and component temperatures at the start-up of natural circulation decay heat removal system. The experimental results showed that natural circulation developed smoothly and the decay heat in the core was removed steadily.

Oral presentation

Development of evaluation methods for decay heat removal by natural circulation under transient conditions, 15; Sodium test for decay heat removal, 2

Kamide, Hideki; Kimura, Nobuyuki; Hayashi, Kenji; Ono, Ayako; Watanabe, Osamu*

no journal, , 

Sodium experiment was carried out for evaluation of natural circulation decay heat removal of Japan sodium cooled reactor, JSFR. Influences of natural circulation in the secondary system and also pressure loss coefficients in the primary and the decay heat removal systems were clarified based on the experiments.

Oral presentation

Sodium experiments on core thermal hydraulics

Kamide, Hideki

no journal, , 

Sodium experiments on core thermal hydraulics carried out in JAEA are introduced as candidates of the next IAEA Coordinated Research Project. Inter-wrapper flow (IWF) is known as natural convection in sodium gap between subassemblies in SFR core due to buoyancy force of cold sodium provided by dipped heat exchanger of DRACS in reactor upper plenum. Sodium experiments on IWF were carried out in this study and thermal hydraulics in the subassemblies and inter-wrapper gaps was clarified.

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