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Kato, Atsushi; Yamamoto, Tomohiko; Ando, Masato; Chikazawa, Yoshitaka; Murakami, Hisatomo*; Oyama, Kazuhiro*; Kaneko, Fumiaki*; Higurashi, Koichi*; Chanteclair, F.*; Chenaud, M.-S.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.11_1 - 11_10, 2022/06
This paper provides an overview of plant system studies to establish a common technical view for Sodium-cooled Fast Reactor concept between France and Japan based on ASTRID600 and the new concept with downsized output called ASTRID150. One of important issues on a reactor structure design is to enhance seismic resistance to be tolerable against strong earthquake such that postulated in Japan. A concept of High Frequency Design is shared, and the design options related to HFD have been examined and design recommendations are established. In addition, this paper include results of studies for a steam generator, a decay heat removal system, a fuel handling system and a containment vessel.
Saito, Hiroyuki*; Yamada, Yumi*; Oyama, Kazuhiro*; Matsunaga, Shoko*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
A self-actuated shutdown system (SASS) is a passive device, which can detach a control rod for reactor shutdown in response to excessive increase in coolant temperature. Since a detachment temperature, which triggers release of a control rod, and a response time are identified as important parameters for validity analyses, this study focused on investigation of the response time and the detachment temperature, and safety analysis to see feasibility of the SASS in low power. For this purpose, design modifications were made to shorten the response time and three-dimensional thermal-hydraulic analysis in a low power operation was carried out in order to confirm the response time. The resulting detachment temperature level is lower than previous studies, leading to improved safety parameters. Based on improved parameters, a safety analysis to see feasibility of the SASS in low power was carried out. From this safety evaluation, it was confirmed that core damage can be prevented by the SASS with flow collector in the case of LOF type ATWS event.
Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*
Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09
Times Cited Count：13 Percentile：73.22(Nuclear Science & Technology)A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.
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
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.
Hagiwara, Hiroyuki; Yamada, Yumi*; Eto, Masao*; Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
The self-actuated shutdown system (SASS), which is selected for Japan Sodium-cooled Fast Reactor (JSFR), is a passive reactor shutdown system utilizing a Curie point electromagnet (CPEM). With CPEM, an excessive fuel outlet temperature rise is sensed and the control rods are released into the core, and the reactor can be shutdown. Therefore it is important for feasibility of SASS to be established by assuring a quick response of CPEM to the coolant temperature rise. In this paper, a device named "flow collector", which collects flows discharged from six fuel subassemblies surrounding CPEM backup control rods, has been proposed to ensure a shorter response time.
Yoshida, Kazuhiro*; Sakata, Hideyuki*; Sago, Hiromi*; Shiraishi, Tadashi*; Oyama, Kazuhiro*; Hagiwara, Hiroyuki*; Yamano, Hidemasa; Yamamoto, Tomohiko
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12
To prevent the vortex cavitations, asymmetric flow in the upper plenum due to the radial slit with upper internal structure (UIS) has been mitigated by installing a cylindrical structure named as dummy plug instead of the fuel handling machine only used for refueling period. In this study, the extended brim and the division plate at the slit of UIS have been proposed in order to improve flow pattern in upper plenum for the purpose of the vortex cavitation prevention.
Eto, Masao*; Kamishima, Yoshio*; Okamura, Shigeki*; Watanabe, Osamu*; Oyama, Kazuhiro*; Negishi, Kazuo; Kotake, Shoji*; Sakamoto, Yoshihiko; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In the JSFR design, the diameter of the Reactor Vessel (RV) shall be minimized and the reactor internal structures shall be simplified for reduction in construction cost. The reduction in the RV diameter is achieved by adopting an advanced refueling system and the hot RV with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the RV is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system.
Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki; Watanabe, Osamu*; Oyama, Kazuhiro*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Design study of an advanced loop-type sodium-cooled fast reactor, JSFR, has been carried out in a frame work of Fast Reactor Cycle Technology Development Project (FaCT) in Japan. As the temperature differences among the control rod channels, blanket assemblies and the core fuel assemblies are 100C centigrade in the maximum, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). In this investigation, a water experiment was conducted using a 1/3 scale 60 sector model of the core and reactor upper plenum. Characteristics of temperature fluctuations near the cold fluid outlets were obtained and it was confirmed that several countermeasures can reduce temperature fluctuations at the bottom of UIS.
Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki; Watanabe, Osamu*; Oyama, Kazuhiro*
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06
An advanced loop type sodium cooled fast reactor, JSFR, has been investigated in the frame work of Fast Reactor Cycle Technology Development Project (FaCT). As the temperatures difference between the control rod channels and the core fuel subassemblies is around 100 C, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). Then, a water experiment was conducted using an 1/3 scale 60 degree sector model. Temperature and its fluctuation intensity distributions around the control rod were measured and an effect of the improved structure against the thermal fatigue was examined.
Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9296_1 - 9296_11, 2009/05
In the present study, three-dimensional thermal-hydraulic analyses of the reactor upper plenum in JSFR were applied to evaluating the following countermeasures including the geometrical structure. The basic ideas of countermeasures are as follows. (1) In order to mitigate the thermal striping phenomenon, locations where hot sodium discharged from fuel assemblies meets with cold sodium from control rods and/or radial blanket subassemblies are kept away from un-replaceable structures above the core. (2) In order to prevent the vortex cavitations, asymmetric flow in the upper plenum due to the radial slit with UIS is mitigated by installing a cylindrical structure named as dummy plug instead of the fuel handling machine only used for refueling period. Furthermore, flow holes on perforated plates with UIS are extended to mitigate radial sodium jet from lower part of UIS. (3) In order to prevent the cover-gas entrainment, Dipped Plate (DP) is installed below the sodium free surface. Original design of DP was a double wall type with many labyrinth seals, so the manufacturing regarded as difficult. Then a single wall type DP has been newly designed and examined in this analyses.
Oyama, Kazuhiro; Kuroha, Takaya*; Takamatsu, Misao; Sekine, Takashi
JNC TN9410 2005-010, 57 Pages, 2005/03
JNC-TN9410-2005-010.pdf:1.51MB
A study of passive safety test using JOYO has been carried out to demonstrate the inherent safety of sodium cooled fast reactors. In this study, emphasis was placed on the improvement of the accuracy of plant kinetics calculations. The Mimir-N2 analysis code, developed to analyze JOYO plant kinetics, was selected as the standard code for predicting plant behavior during transients.Mimir-N2 was previously modified for MK-III core in 2001, 2002. Then, it implemented MK-III performance testing estimate analysis. The MK-III performance test included manual reactor shutdown test and loss of power supply test etc. as transient tests. In order to further improve the accuracy of the calculation, the Mimir-N2 heat transport system models of the reactor vessel upper plenum, the hot leg of secondary heat transport system and the dump heat exchanger were modified based on the results of the MK-III performance test in 2003.In this year, it stores up to get the prospect to have paid to the implementation of the UTOP, the ULOF test which is planned as the passive safety test, it evaluated about the plant structure and UTOP, the ULOF analysis for the parameter of which was the investing reactivity and so on by the Mimir-N2 analysis code. As a result, we could work out the testing condition which has prospect.
Oyama, Kazuhiro; Kawahara, Hirotaka; Ishida, Koichi; Ariyoshi, Masahiko; Isozaki, Kazunori; Sugaya, Kazushi*; Fukami, Akihiro*
JNC TN9410 2005-006, 121 Pages, 2005/03
JNC-TN9410-2005-006.pdf:10.81MB
In the MK-III performance test, the experimental fast reactor JOYO raised the reactor thermal power gradually with about 20%, 50%, 75%, 90%, and 100% (140MWt), and reached 140MWt which are the full power of a MK-III reactor core on October 28, 2003. Then, continuation operation beyond full power 100 hour was attained. This report summarized the result of power-up test , full power continuation operation test, blower start-up test.The outline is as follows.(1)From the standby state (system temperature of 250degree), the usual power-up operation (an power-up rate ;about 5MWt/20min, a power is held for about 10 minutes every 5MWt) attained the reactor thermal full power (140MWt) gradually on October 28,2004. Moreover, it checked that each part temperature and flow were less than alarm setting values on each power level.(2)The reactor thermal power was made into the parameter, a series of operations about the blower start-up, and the influence which it has on coolant temperature was checked. As a result, the optimal reactor thermal power which starts up the blower from a natural ventilation cooling state was set to about 18 MWt, and the starting procedure was made into the method(order of 1A-2A-1B-2B) which starts four sets of the one blower at a time one by one.(3)It checked that reactor shutdown operation by two control-rod simultaneous insertion at 35MWt, and it could carry out with time margin with a series of sufficient operations of resulting from control rod insertion in the blower shutdown. By adopting this reactor shutdown operation method, operation of an operation stuff was mitigated and it checked that plant characteristics also improved.(4)The reactor full power was reached on November 14. Continuation operation beyond full power 100 hour was attained after that till on November 20, 10:30. The data of each part of a plant was acquired at intervals of 24 hours, and it checked that it was less than an alarm setting value.
Oyama, Kazuhiro; Kawahara, Hirotaka; Ariyoshi, Masahiko; Isozaki, Kazunori; Sugaya, Kazushi*; Fukami, Akihiro*
JNC TN9410 2005-005, 56 Pages, 2005/03
JNC-TN9410-2005-005.pdf:14.56MB
The experimental fast reactor JOYO MK-III increased that the reactor thermal power by the factor 1.4. The main intermediate heat exchangers (IHX) and the dump heat exchangers (DHX) were exchanged. And then, the flow rate of the main cooling system, the secondary cooling system were increased. As one of the performance test to confirm that the cooling system which included these switch receptacles has an enough decay heat performance, it did an heat transfer characteristics test and it evaluated a heat balance, the decay heat performance of IHX and DHX.The outline is as follows.(1)It confirmed that the modificated plant had fixed performance by the heat balance of full power.(2)The secondary inlet temperature of B-loop IHX is about 6degree higher out of the cooling system with A-loop. It thinks that this is one because of the difference ( about 2 % ) with the flow rate of the main cooling system in measurement. There was decay heat capacity of the A-loop and the B-loop in the balance, making the flow rate of the main cooling system of the A-loop positive and supposing that the B-loop is a revision flow rate and as for the heat transfer performance of IHX of the A-loop and the B-loop, the approximately equal thing could be confirmed. As a result, as for the overall heat transfer coefficient of IHX, the A-loop was about 125 % of the design value, the B-loop was about 129 % of the design value and it confirmed that two IHX had the performance to be equal and an enough decay heat performance.(3)It made DHX outlet air temperature about 20degree and it calculated DHX outlet air flow from the decay heat capacity from sodium coolant and the DHX outlet air temperature in full power. As a result, DHX could confirm that the decay heat ability to be equivalent to he reactor thermal power in 85 - 90 % of capacities of design value (6,750m
/min) and an enough decay heat performance.
Oyama, Kazuhiro; Kawahara, Hirotaka; Ariyoshi, Masahiko; Sugaya, Kazushi*; Fukami, Akihiro*
JNC TN9410 2005-004, 74 Pages, 2005/03
JNC-TN9410-2005-004.pdf:14.44MB
In the MK-III performance test, the experimental fast reactor JOYO measured the reactor thermal power in each step from a low power to a full power, and calibrated the intermediate range neutron monitors and power range neutron monitors of core instrumentation equipment. This report summarized the result of thermal power calibration.The outline is as follows. (1) We measured the reactor thermal power in each step from a low power to a full power, and calibrated the intermediate range neutron monitors and power range neutron monitors of core instrumentation equipment. Between the power range neutron monitors and the reactor thermal power, it has confirmed from this that there was good linearity. (2) From transition of November 20 [November 14 to] reactor thermal power, and the graphite temperature, although the graphite temperature (83-5,6,7) is rising to about 97 degree, it is mostly saturated in the 6th day after the full power attainment. In addition, thermal power calibration was carried out 4 times within this period. (November 14th, 15th, 16th, 18th) (3) All the errors of the full power measurement at the time of rating are 
3.42% (=4.8 MWt). It is in the reactor thermal power error used by the thermal design of a MK-III reactor core (3.6%). (4) The compensation coefficient in
Ito, Chikara; Ito, Hideaki; Ishida, Koichi; Hatoori, Kazuhiro; Oyama, Kazuhiro; Sukegawa, Kazuya*; Murakami, Takanori; Kaito, Yasuaki; Nishino, Kazunari; Aoyama, Takafumi; et al.
JNC TN9410 2005-003, 165 Pages, 2005/03
JNC-TN9410-2005-003.pdf:12.66MB
At experimental fast reactor JOYO, appraisal of detection efficiency of behavior and FFD and FFDL of the fission product which is discharged inside the furnace as one of safety research of the country, is carried out. In MK-II core, the slit in the gas plenum part of the test sub-assembly, the test which irradiates this(1985 April, FFDL in-pile test(I)), providing the slit in the fuel column part of the test sub-assembly, the test which it irradiates(1992 November, FFDL in-pile test(II)) were carried out.MK-III reactor core replacement was completed and started in 2004. That the behavior in the system of FP with the reactor core replacement and so on changes in the MK-III reactor core and to have an influence on the sensitivity and the replying of FFD and FFDL are thought of. Therefore, behavior of FP in the fuel failure in the MK-III reactor core, the performance of FFD and FFDL must be confirmed beforehand. Moreover, to prepare for the fuel failure and the RTCB test which is doing a future plan, and to confirm a plant operation procedure in the fuel failure in MK-III reactor core operation and to attempt for the correspondency to improve are important.Therefore, in the period from 2004 November 11th to November 29th, it carried out the FFDL in-pile(III). It did a series of plant operation to stop a nuclear reactor after loading a reactor core center with the fuel element for the test which provided an artificial slit for the fuel cladding in the MK-III reactor core and irradiating it and detecting fuel damaging and to take out fuel. And it confirmed the operation procedure of the fast reactor in the fuel failure.Also, the improvement items such as the improvement of the operation and the procedure and the remodeling and the service of the facilities could be picked up. In the future, it attempts these compatible, and it prepares for the MK-III reactor core operation and it incorporates a final examination result by the improvement of the safety of FBR.
Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro; Sekine, Takashi; Ariyoshi, Masahiko; Ito, Chikara; Masaaki, Nemoto; Murakami, Takanori; Isozaki, Kazunori; Hoshiba, Hideaki; et al.
JNC TN9410 2003-011, 197 Pages, 2004/03
JNC-TN9410-2003-011.pdf:10.26MB
MK-III performance tests began in June 2003 to fully characterize the upgraded core and heat transfer system. Then, the last pre-use inspection was finished in November 2003.This report summarize the result of each performance test.
Oyama, Kazuhiro; Aoyama, Takafumi
JNC TN9200 2003-005, 175 Pages, 2004/03
JNC-TN9200-2003-005.pdf:17.64MB
On 2004 February 26th it hosted the conference which does a communication about the current status and the technical development,etc of JOYO , CEFR and MONJU.Through the meeting, a discussion was done for the information exchange meeting and the mutual sending of the technical expert as the cooperation of the future. We were possible to confirm about the field to be interested in CEFR.
Sakai, Takaaki; ; *; Oyama, Kazuhiro*
JNC TN9400 2001-052, 71 Pages, 2001/05
JNC-TN9400-2001-052.pdf:3.24MB
The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with a11 the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300
550MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxilialy Cooling System (PRACS) in the 400MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate ...
Oyama, Kazuhiro*; Watanabe, Osamu*; *
JNC TJ9410 2001-002, 93 Pages, 2000/03
In the Strategic Research to Commercialize Fast Breeder Reactor Cycle plan, various breeder reactor core concepts are studied which are not restricted to the MOX-sodium combination. Metal and nitride are studied for fuels and gas, water, and lead for coolants. The objectives of this study is to compare the safety characteristics of the various breeder reactor cores by assuming the situation of the post-accident heat removal after hypothetical core disruptive accident. As a preliminary evaluation, coolable limit of core debris beds, which are formed after hypothetically disrupted core, was evaluated for the combinations of three types of fuels, MOX, metal and nitride, and four types of coolants, liquid sodium, lead, water and carbon dioxide gas. For the evaluation, a one-dimensional version of the DEBRIS-MD code which models the temperature distribution in a debris bed was used. Although the original code can handle only sodium coolant, special versions have been developed to handle lead, water and carbon dioxide gas coolants. Furthermore, the computer code for calculating debris bed temperature distribution was integrated in a newly developed coolant flow calculation model. It can handle arbitrary combination of coolant flow paths by using one dimensional flow network modeling. The computer code, named DEBNET was successfully used to analyze the post-accident heat removal in a 600MWe class FBR plant.
Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
no journal, ,
Design validity evaluation was carried out, which was focused on thermal striping, gas entrainment and vortex cavitations included in Thermal-Hydraulic issues in the reactor upper sodium plenum that is important to establish the outline design of Large-Sized Sodium-Cooled Fast Reactor.
