Ichihara, Yoshitaka*; Nakamura, Naohiro*; Nabeshima, Kunihiko*; Choi, B.; Nishida, Akemi
Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 10 Pages, 2022/07
The objective of this study is to evaluate the applicability of the equivalent linear analysis method for reinforced concrete, which uses frequency-independent complex damping with a small computational load, to the seismic design of reactor building of the nuclear power plant. To achieve this, the three-dimensional finite element analyses of the soil-structure interaction system focusing on the nonlinear and equivalent linear seismic behavior under an ideal soil condition were performed for Kashiwazaki-Kariwa nuclear power plant Unit 7 reactor building. From these results, the equivalent linear analysis method showed a generally good correspondence with the nonlinear analysis method, and the effectiveness of the method was confirmed.
Ichihara, Yoshitaka*; Nakamura, Naohiro*; Nabeshima, Kunihiko*; Choi, B.; Nishida, Akemi
Kozo Kogaku Rombunshu, B, 68B, p.271 - 283, 2022/04
This paper aims to evaluate the applicability of the equivalent linear analysis method for reinforced concrete, which uses frequency-independent hysteretic damping, to the seismic design of reactor building of the nuclear power plant. To achieve this, we performed three-dimensional FEM analyses of the soil-structure interaction system, focusing on the nonlinear and equivalent linear seismic behavior of a reactor building under an ideal soil condition. From these results, the method of equivalent analysis showed generally good correspondence with the method of the nonlinear analysis, confirming the effectiveness. Moreover, the method tended to lower the structural stiffness compared to the nonlinear analysis model. Therefore, in the evaluation of the maximum shear strain, we consider that the results were more likely to be higher than the results of nonlinear analysis.
Subekti, M.*; Kudo, Kazuhiko*; Nabeshima, Kunihiko; Takamatsu, Kuniyoshi
Atom Indonesia, 43(2), p.93 - 102, 2017/08
Reactor kinetics based on point kinetic model have been generally applied as the standard method for neutronics codes. As the central control rod (C-CR) withdrawal test has demonstrated in a prismatic core of HTTR, the transient calculation of kinetic parameter, such as reactivity and neutron fluxes, requires a new method to shorten calculation-process time. Development of neural network method was applied to point kinetic model as the necessity of real-time calculation that could work in parallel with the digital reactivity meter. The combination of TDNN and Jordan RNN, such as TD-Jordan RNN, was the result of the modeling approach. The application of TD-Jordan RNN with adequate learning, tested offline, determined results accurately even when signal inputs were noisy. Furthermore, the preprocessing for neural network input utilized noise reduction as one of the equations to transform two of twelve time-delayed inputs into power corrected inputs.
Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04
A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.
Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/06
Under natural circulation decay heat removal conditions, three characteristic phenomena; flow redistribution in the core as well as in the fuel subassemblies, inter-subassembly heat transfer and gap flow between wrapper tubes of fuel subassemblies are important for assessing the temperature distribution in the core. In order to improve the prediction accuracy, a whole core model which can consider these three phenomena has been incorporated into the plant dynamics analysis code Super-COPD. In this study, analyses of two kinds of sodium experiments were performed to validate Super-COPD with the whole core model, which were focusing on inter-subassembly heat transfer phenomena.
Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08
Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.
Kato, Atsushi; Chikazawa, Yoshitaka; Nabeshima, Kunihiko; Iwasaki, Mikinori*; Akiyama, Yo*; Oya, Takeaki*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.593 - 600, 2015/05
Japan sodium cooled fast reactor is the advanced loop type reactor developing in Japan. After the Fukushima-Dai-ichi NPP accident, system enhancement against severe accident have been investigated mainly for residual decay heat removal system, spent fuel storage system and emergency power sources in order to satisfy the safety design criteria for Generation IV SFR. This paper describes principle of the building layout design and the actual approach to be consistent with the recent design enhancement in JSFR. From the perspective of greater ability to withstand severe events, the principles of the building layout design as the measures against aircraft attack and the consequential fire, and tsunami are introduced in order to avoid local event initiating and simultaneous redundant failure of the safety grade facilities and could achieve lowering risk of the loss of all stuck and maintaining the essential power supply.
Chikazawa, Yoshitaka; Kato, Atsushi; Nabeshima, Kunihiko; Otaka, Masahiko; Uzawa, Masayuki*; Ikari, Risako*; Iwasaki, Mikinori*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Design study and evaluation for SDC and safety SDG on the BOP of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system and air conditioning, requirements and relation between safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.
Nabeshima, Kunihiko; Aizawa, Kosuke; Chikazawa, Yoshitaka; Sato, Daisuke*; Ikari, Risako*
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.600 - 606, 2014/04
One of the advantages in features of Japan Sodium-cooled Fast Reactor (JSFR) is that it's free from high-capacity power load or quick activation of emergency power supply because Decay Heat Removal System (DHRS) adopted natural circulation. However, the Emergency Power Supply System of JSFR is reconsidered to improve the reliability for station blackout (SBO) caused by extreme external events after the nuclear accident at Fukushima. A non-class 1E GTG are added for practically elimination of Loss of Heat Removal System (LOHRS), and alternative power supply system composed of small air-cooled DG and lead-acid batteries is newly introduced to maintain class 1E equipment against the long-term station blackout (SBO).
Kugo, Teruhiko; Akie, Hiroshi; Yamaji, Akifumi; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9371_1 - 9371_8, 2009/05
Combining a nuclear reactor with thermoelectric converters is expected to be one of promising options to supply a propulsion power for deep space explorers. One of the key features of the concept is to use low enriched uranium fuels from the viewpoint of nuclear non-proliferation. Fuels of uranium oxide, nitride and metal were examined. Zirconium and yttrium hydrides, beryllium, zirconium beryllide and graphite were considered as moderators. Reflectors of beryllium, beryllium oxide, zirconium beryllide and graphite were taken into consideration. A criticality survey of the core was performed by changing the ratio of the fuel, moderator and structure, and the reflector thickness. As a result from the viewpoint of a smaller mass of reactor, it is better to use thermal spectrum cores than fast ones, and the metal hydride moderators than beryllium or graphite. For example, a combination of uranium nitride, yttrium hydride and beryllium reflector achieves a reactor mass of as low as 500kg.
Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9366_1 - 9366_8, 2009/05
This study has been carried out in series with the other study, "Criticality of Low Enriched Uranium Fueled Core" to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of three different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The three systems are the solid thermal conduction system (STC), core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and cover down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept.
Yoshida, Hiroyuki; Suzuki, Takayuki*; Nabeshima, Kunihiko; Takase, Kazuyuki
Nihon Kikai Gakkai Kanto Shibu Dai-15-Ki Sokai Koenkai Koen Rombunshu, p.105 - 106, 2009/03
no abstracts in English
Shimizu, Atsushi; Nabeshima, Kunihiko; Nakagawa, Shigeaki
JAEA-Technology 2008-082, 44 Pages, 2009/01
The High temperature engineering test reactor (HTTR) executed the rated power driving for 30 days of the first time (850C in temperature of the nuclear reactor exit coolant) until March, 27th through April, 26th, 2007. In this operation, HTTR was observed according to the operation monitoring model with the neural network, and the detection performance of neural network was verified during slight changes of reactor state at rated power. The neural network used for the operation monitoring was an auto-associative network, where 31 input 31 outputs and the hidden layers were connected with 20 units by the hierarchy of three layer structure. Back-propagation algorithm is used for study rule. The operation monitoring model in initial study was constructed by using the power up data between 30% and rated power, which are randomly studied. The adjustment study during the operation monitoring changes the internal structure of the initial study model to follow the changes of reactor status, such as the combustion of the nuclear fuel for the rated power driving. As a monitoring result, slight changes of reactor state by the control system operation were correctly detected, and the on-line application to an early anomaly diagnosis for HTTR facilities will be expected.
Nabeshima, Kunihiko; Subekti, M.*; Matsuishi, Tomomi*; Ono, Tomio*; Kudo, Kazuhiko*; Nakagawa, Shigeaki
Journal of Power and Energy Systems (Internet), 2(1), p.92 - 103, 2008/00
The neural networks have been utilized in on-line monitoring-system of High Temperature Engineering Tested Reactor (HTTR) with thermal power of 30 MW. In this system, several neural networks can independently model the plant dynamics with different architecture, input and output signals and learning algorithm. One of main task is real-time plant monitoring by Multi-Layer Perceptron (MLP) in auto-associative mode, which can model and estimate the whole plant dynamics by training normal operational data only. Other tasks are on-line reactivity prediction, reactivity and helium leak monitoring, respectively. From the on-line monitoring results at the safety demonstration tests, each neural network shows good prediction and reliable detection performances.
Subekti, M.*; Kudo, Kazuhiko*; Nabeshima, Kunihiko
Proceedings of International Conference on Advances in Nuclear Science and Engineering (ICANSE 2007) (CD-ROM), p.53 - 63, 2007/11
The monitoring system for a huge and complex Pressurized Water Reactor (PWR) has some difficulties of monitoring task due to the dynamic system with large number of plant signals. The anomaly detection using neural networks integrated in the monitoring system improves the reactor safety to detect the anomaly faster than conventional methods. The advanced research considered the online anomaly diagnosis using expert system to complete the monitoring system tasks. The combination of neural network and expert system (neuro-expert) has been developed and tested in some anomaly conditions using PWR simulator. In simulation, the neuro-expert system could detect and diagnose the anomalies faster than the conventional alarm system.
Nabeshima, Kunihiko; Nakagawa, Shigeaki; Makino, Jun*; Kudo, Kazuhiko*
Proceedings of International Symposium on Symbiotic Nuclear Power Systems for 21st Century (ISSNP) (CD-ROM), p.142 - 147, 2007/07
Two types of neural networks have been utilized for real-time condition monitoring of High Temperature Engineering Tested Reactor (HTTR) in JAEA, Japan. Multi-Layer Perceptron (MLP) in auto-associative mode could model the whole plant dynamics and detect many kind of abnormal conditions. Another neural network with feedback connection can estimate the occurrence time and amount of helium leakage after auto-associative MLP detects the anomaly.
Nabeshima, Kunihiko; Matsuishi, Tomomi*; Makino, Jun*; Subekti, M.*; Ono, Tomio*; Kudo, Kazuhiko*; Nakagawa, Shigeaki
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
The neural networks have been utilized in on-line monitoring system of High Temperature Engineering Tested Reactor (HTTR) with thermal power of 30MW. In this system, several neural networks can independently model the plant dynamics with different architecture, input and output signals and learning algorithm. One of main task is real-time plant monitoring by Multi-Layer Perceptron (MLP) in auto-associative mode, which can model and estimate the whole plant dynamics by training normal operational data only. Other tasks are on-line reactivity prediction, reactivity and helium leak monitoring, respectively. From the on-line test results, each neural network shows good prediction and reliable detection performances.
Subekti, M.*; Ono, Tomio*; Kudo, Kazuhiko*; Nabeshima, Kunihiko; Takamatsu, Kuniyoshi
Proceedings of 5th American Nuclear Society International Topical Meeting on Nuclear Plant Instrumentation, Controls, and Human Machine Interface Technology (NPIC & HMIT 2006) (CD-ROM), p.75 - 82, 2006/11
The neuro-expert has been utilized in previous monitoring-system research of Pressure Water Reactor (PWR). The research improved the monitoring system by utilizing neuro-expert, conventional noise analysis and modified neural networks for capability extension. The parallel method applications required distributed architecture of computernetwork for performing real-time tasks. The research aimed to improve the previous monitoring system, which could detect sensor degradation, and to perform the monitoring demonstration in High Temperature Engineering Tested Reactor (HTTR). The developing monitoring system based on some methods that have been tested using the data from online PWR simulator, as well as RSG-GAS (30 MW research reactor in Indonesia), will be applied in HTTR for more complex monitoring.
Nabeshima, Kunihiko; Kurnianto, K.*; Surbakti, T.*; Pinem, Surian*; Subekti, M.*; Minakuchi, Yusuke*; Kudo, Kazuhiko*
Proceedings of ICSC Congress on Computational Intelligence Methods and Applications (CIMA'2005) (CD-ROM), 4 Pages, 2005/12
The ANNOMA (Artificial Neural Network of Monitoring Aids) system is applied to the condition monitoring and signal validation of Multi Purpose Reactor in Indonesia. The feedforward neural network in auto-associative mode learns reactor's normal operational data, and models the reactor dynamics during the initial learning. The basic principle of the anomaly detection is to monitor the deviation between the process signals measured from the actual reactor and the corresponding values predicted by the reactor model, i.e., the neural networks. The pattern of the deviation at each signal is utilized for the identification of anomaly, e.g. sensor failure or system fault. The on-line test results showed that the neural network successfully monitored the reactor status during power increasing and steady state operation in real-time.
Subekti, M.*; Ono, Tomio*; Kudo, Kazuhiko*; Takamatsu, Kuniyoshi; Nabeshima, Kunihiko
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
In this study, a new full integrated monitoring system scheme based on distributed architecture is proposed. This monitoring system has a distributed architecture; monitoring tasks are assigned to client PCs by the central server. As a result of the distributed architecture, it is expected that the processing capabilities is maximized and the real time consistency is not impaired even if heavy monitoring tasks cause a shortage of bandwidth. And this system integrates signal processing modules in the main system and the main system distributes the monitoring tasks on its client PCs with TCP-IP technology. Signal processing between the main system and the client PCs is optimized so that monitoring tasks are distributed very efficiently. And, each client PC is completely separated, processing condition of one PC never effects on the other PC's processing.