Kudo, Hideyuki*; Otani, Yuichi*; Hara, Masahide*; Kato, Atsushi; Otaka, Masahiko; Ide, Akihiro*
Journal of Nuclear Science and Technology, 57(4), p.408 - 420, 2020/04
In a fuel handling system of sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP) in order to minimize plant operating loads. A next-generation SFR in Japan has adopted an advanced dry cleaning process which consists of the following steps, argon gas blowing to remove the metallic residual sodium on the FA, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP. This three-step process increases economic competitiveness and reduces waste products thanks to a waterless process. In this R&D work, performance of the dry cleaning process has been investigated.
Abe, Yuta; Otaka, Masahiko; Okazaki, Kodai*; Kawakami, Tomohiko*; Nakagiri, Toshio
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 7 Pages, 2019/05
Since the hardness of fuel debris containing boride from BC pellet in control rod is estimated to be two times higher as that of oxide, such as UO and ZrO, it is necessary to select the efficient and appropriate operation for removal of fuel debris formed in the severe accident of nuclear power plants. We focused on the characteristics of LIBS, an innovative rapid chemical in-situ analysis technology that enables simultaneous detection of B, O, and other metal elements in fuel debris. Simulated solidified melt specimens were obtained in the plasma heating tests (CMMR-0/-2, performed by JAEA) of simulated fuel assembly (ZrO is used to simulated UO pellet, other materials such as stainless steel, BC are same as fuel assembly). The LIBS signals of (B/O)/Zr ratio showed good linear relationship with Vickers hardness. This technique can be also applied as in-situ assessment tool for elemental composition and Vickers hardness of metal-oxide-boride materials.
Kudo, Hideyuki*; Otani, Yuichi*; Hara, Masahide*; Kato, Atsushi; Ishikawa, Nobuyuki; Otaka, Masahiko; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki; Ide, Akihiro*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05
A next generation SFR in Japan has adopted an advanced dry cleaning system which consists of the argon gas blowing process to reduce the amount of metallic residual sodium remaining on spent fuel assemblies. This paper describes experimental and analytical work focusing on the amount of residual sodium remaining on a fuel pin bundle before and after the argon gas blowing process. The experiments were conducted using a sodium test loop and a short specimen consisting of a 7 pin bundle. The effects of the blowing gas velocity and the blowing time were quantitatively analyzed in the experiments. On the basis of these experimental results, evaluation models predicting the amount of the residual sodium were constructed.
Ide, Akihiro*; Kudo, Hideyuki*; Inuzuka, Taisuke*; Hara, Masahide*; Kato, Atsushi; Ishikawa, Nobuyuki; Otaka, Masahiko; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05
A next generation SFR in Japan has adopted an advanced dry cleaning system which consists of the following process of argon gas blowing to reduce the amount of metallic sodium, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP without using storage containers. This three-step process increases economic competitiveness and reduces waste products. In this Research and Development work, the amount of residual sodium and performance of the dry cleaning process were investigated. This paper describes experimental and analytical work for all parts of a fuel assembly except for a fuel pin bundle.
Kotake, Shoji*; Chikazawa, Yoshitaka; Takaya, Shigeru; Otaka, Masahiko; Kubo, Shigenobu; Arai, Masanobu; Kunogi, Kosuke; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
A maintenance management required to prototype nuclear power reactors is proposed. Monitoring and control of sodium impurity and thermal transient are extremely important for sodium boundary maintenance for sodium-cooled fast reactors. At the fast stage of the prototype reactor Monju operation, degradation mechanism on the piping should be demonstrated based on operation experiences. Therefore inspection on a representative position for crack indication and pipe thickness is proposed. Due to less experience of SFR plants, early detection of boundary failure is considered. For a matured operation stage, when degradation mechanism is well demonstrated based on inspection data, inspection cycle could be extended. And for commercial reactors, maintenance without inspection will be established based on accumulated operation experiences including those of the prototype reactor Monju.
Kato, Atsushi; Nagai, Keiichi; Ara, Kuniaki; Otaka, Masahiko; Oka, Nobuki*; Tanaka, Masako*; Otani, Yuichi*; Ide, Akihiro*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
In a fuel handling system (FHS) of a sodium-cooled fast reactor, it is necessary to reduce residual sodium on a spent fuel subassembly before storing at a spent fuel water-pool (SFP) in order to minimize design loads. Although the wet cleaning process adopted on MONJU could eliminate almost all of residual sodium, a large amount of radioactive liquid waste occurs and it needs long duration of cleaning treatment and large plant commodities. On the other hand, Japan sodium-cooled fast reactor adopted an advanced dry cleaning system which consists of roughly blowing massive sodium on the fuel subassembly out by 300C argon gas, inactivation of residual sodium to NaOH by moist argon gas and directly immersion into the SFP to achieve economic competitiveness and waste reduction. This paper reports current status of recent R&D activities to demonstrate a performance of the dry cleaning process in Japan which are for improvement of the cleaning performance and optimizing the FHS design.
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.
Furukawa, Tomohiro; Hirakawa, Yasushi; Kato, Shoichi; Iijima, Minoru; Otaka, Masahiko; Kondo, Hiroo; Kanemura, Takuji; Wakai, Eiichi
Fusion Engineering and Design, 89(12), p.2902 - 2909, 2014/12
For the irradiation test of the candidate materials for the fusion DEMO reactor, Engineering Validation and Engineering Design Activities (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) are performed under the Broader Approach Activities. As a major Japanese activity on the target facility of the IFMIF, the engineering validation using the EVEDA Lithium Test Loop which is the largest scale liquid lithium test loop has been started in 2012. In parallel with the design and fabrication, the research on the technology establishment for the lithium safety handling was started in 2008, as one of the related technologies under the IFMIF-EVEDA. In the research, experiments of lithium chemical reaction, experiments on lithium fire, establishment of chemical analysis of impurities in lithium and experiments on advanced lithium leak detection system were carried out. This paper describes the results of these experiments.
Furukawa, Tomohiro; Otaka, Masahiko; Hirakawa, Yasushi
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 5 Pages, 2014/07
This study which consists of two kinds of experiments has been performed as one of the related technologies under the IFMIF-EVEDA. The one is the experiments on advanced lithium leak detection system, and the other is the experiments on fire-extinguishing behavior of lithium. In the experiments on advanced lithium leak detection system, the potential of the small leak detection system using a laser-induced breakdown spectroscopy has been investigated. On the fire-extinguishing behavior of lithium, fire-extinguishing performance of two kinds of carbon based extinguishants (MITEX and GRAPHEX) was investigated in the experiments.
Otaka, Masahiko; Kato, Atsushi; Chikazawa, Yoshitaka; Uzawa, Masayuki*; Ide, Akihiro*; Kaneko, Fumiaki*; Hara, Hiroyuki*
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.607 - 615, 2014/04
Responding to the TEPCO's Fukushima Dai-ichi Nuclear Power Station accident, the design study was carried out to enhance cooling function of fuel handling system for Japan sodium-cooled fast reactor (JSFR). Design measures to maintain an FHS cooling function even in case of Design Extension Conditions are studied; such as alternative cooling systems which are independent and diversified from design basis cooling systems. Based on effectiveness evaluations, it was confirmed that these measures have adequate availability against the postulated design extension conditions.
Nakamura, Kazuyuki; Furukawa, Tomohiro; Hirakawa, Yasushi; Kanemura, Takuji; Kondo, Hiroo; Ida, Mizuho; Niitsuma, Shigeto; Otaka, Masahiko; Watanabe, Kazuyoshi; Horiike, Hiroshi*; et al.
Fusion Engineering and Design, 86(9-11), p.2491 - 2494, 2011/10
In IFMIF/EVEDA, tasks for lithium target system are shared to 5 validation tasks (LF1-5) and a design task (LF6). The purpose of LF1 task is to construct and operate the EVEDA lithium test loop, and JAEA has a main responsibility to the performance of the Li test loop. LF2 is a task for the diagnostics of the Li test loop and IFMIF design. Basic research for the diagnostics equipment has been completed, and the construction for the Li test loop will be finished before March in 2011. LF4 is a task for the purification systems with nitrogen and hydrogen. Basic research for the purification equipment has been completed, and the construction of the nitrogen system for the Li test loop will be finished before March in 2011. LF5 is a task for the remote handling system with the target assembly. JAEA has an idea to use the laser beam for cutting and welding of the lip part of the flanges. LF6 is a task for the design of the IFMIF based on the validation experiments of LF1-5.
Hashimoto, Makoto; Otaka, Masahiko; Ara, Kuniaki; Kanno, Ikuo; Imamura, Ryo*; Mikami, Kenta*; Nomiya, Seiichiro*; Onabe, Hideaki*
Journal of Nuclear Science and Technology, 46(1), p.76 - 82, 2009/01
A photon detector for unfolding X-ray energy distribution has been previously reported. Simulation studies on the unfolding method for this detector are discussed for diagnostic X-rays and mixed-source rays of Cs/Co. Response functions for diverse diagnostic X-rays are almost indistinguishable and an unfolding using error reduction method is not sufficiently worked. The spectrum surveillance method is effective in this case. However, a simple error reduction method is useful for mixed rays. For both techniques, the neural network method is promising.
Nippon Kikai Gakkai Doryoku Enerugi Shisutemu Bumon Nyusu Reta, (32), p.5 - 6, 2006/05
no abstracts in English
Otaka, Masahiko; Hirabayashi, Masaru; Ara, Kuniaki
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), P. (50439), 2005/05
In order to enhance the maintenance performance of Sodium-Cooled Fast Reactor (SFR), a novel in-sodium imaging technique was proposed for monitoring a deformation or a displacement of in-sodium components or assemblies in primary cooling systems. This technique based on computed tomography using high-energy gamma photons emitted from radioisotopes of sodium (22Na, 24Na) in the primary coolant. Feasibility studies have been performed in order to explore the applicability to typical piping (or vessels). Preliminary analyses exploring measuring time and imaging capability were performed. The results showed that the proposed technique was feasible as a monitoring system of in-sodium components.
Nagai, Keiichi; Nagai, Keiichi; Otaka, Masahiko; Miyakoshi, Hiroyuki; Onojima, Takamitsu
JNC-TN9400 2003-058, 35 Pages, 2003/05
A preliminary examination was carried out for evaluation of the detection sensitivity of Laser Sodium Leak Detector (LLD) based on a principle of Laser Induced Breakdown Spectroscopy (LIBS). Evaluation criteria and examination conditions were planned based on the results of preliminary experiments.The main results are as follows:(1) Signal intensity of LLD was obtained with parameter of sodium concentration in combustion aerosols. The signal intensity in the combustion aerosols was nearly equivalent to that in case of sodium mist using carrier gas of nitrogen. It was shown that LLD was effective to detect sodium in the combustion aerosols.(2) Diameter or chemical component of sodium aerosols are one of significant factors for the detection sensitivity of LLD. Preliminary experiments were carried out with parameters of humidity, oxygen concentration, and pressure of carrier gas. The obtained experimental data of a few cases showed that influence of these parameters was limited on the detection sensitivity of LLD.3) Based on the preliminary experimental results, main conditions of a sensitivity evaluation test plan were decided for LLD.
Hirabayashi, Masaru; Otaka, Masahiko; Hayashida, Hitoshi; Ara, Kuniaki
JNC-TN9400 2003-016, 35 Pages, 2003/04
To confirm structural integrity of a primary cooling system and in-vessel components in a sodium-cooled fast breeder reactor, monitoring and inspection technique applying Gamma-rays emitted from sodium are proposed. The basic principle is as follows. As radioisotope Na decays, photons are emitted and a fraction of these photons penetrate materials. If the number of these photons is counted by radiation detectors, an image of gamma-rays source is reconstructed by a computed tomography technique. In this report, Applicability and problems concerned with the technique are investigated. Main results are as follows: (1)To verify applicability, the technique was analytically investigated based on gamma-rays emitted from sodium coolant in a typical pipe of a primary cooling system. As a result, it was confirmed that the image of gamma-rays source could be reconstructed. (2)A required time to measure in a spatial resolution of about 1mm was investigated in the detection efficiency of 20%. The time was about 4 minutes per section by a thousand detectors in the typical pipe of a primary cooling system. And in a typical steam generator, the time was about 2 days per section by ten thousand detectors. (3)To realize a fluoroscopic inspection system, it is necessary that the principle should be verified by experimental researches. Main equipments of the system are a collimator, radiation detector, scanner, signal processing device and image processing device. As a spatial resolution is decide by the collimator, the shape must be evaluated by experimental researches and analytical investigation.
Otaka, Masahiko; Ohshima, Hiroyuki
PNC-TN9410 97-091, 37 Pages, 1997/10
Whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors under various operation conditions, e.g., normal operation and transition from forced to natural circulation. This second report describes the core model and its verification analysis. ACT consists of several thermal-hydraulic calculation modules related to the following regions: (1)subassemblies (fuel, blanket and control rods) (2)inter-subassembly gaps, (3)upper plenum and (4)primary heat transport system. The subassembly module is almost equivalent to the subchannel analysis code ASFRE-III, which was developed and validated at Thermal Hydraulic Research Section, and is applied to each subassembly of the core. The inter-subassembly gap module is used for calculating the flow and temperature fields in the gaps between the wrapper tubes (inter-wrapper flow) and was also developed based on ASFRE-IIl code (the first report described last year). The upper plenum and primary heat transport system modules are utilized to offer complicated boundary conditions to the whole-core analysis especially under natural circulation conditions with the operation of the direct reactor auxiliary coolant system. In this work, the core model as the main part of ACT was developed by coupling the subassembly modules and the inter-subassembly gap module and it has made possible to calculate flow and temperature fields in the whole core including thermal interaction between the inner subassemblies and the inter-subassembly gaps. The coupling was made explicitly through the heat exchange on the wrapper tube outer surface from the viewpoints of flexibility to model various core geometry and program parallelization for large-scale simulation. The core model was applied to the analysis of PLANDTL-DHX sodium experiment whose test section consisted of 7 subassemblies for code verification. It was confirmed that the predicted sodium ...
Kanno, Ikuo*; Uesaka, Akio*; Otaka, Masahiko; Hashimoto, Makoto; Ara, Kuniaki; Nomiya, Seiichiro*; Onabe, Hideaki*
no journal, ,
The beam hardening free CT imaging, which is not effected with the size of speciman, was obtained with the energy and the energy differential of X-rays. However, measurement of X-ray energy in high density as in CT is hard. To solve this problem, current mode X-ray detector (CMD) was designed. CMD is configured with several detective elements aligned with X-ray fluence, and read the current from each detective element. Relatively low energy photon is mainly absorbed in front detective elements of CMD and high energy photon can reach to deep, rough X-ray spectrum can be expected with current from each element. The ability of CMD is examined by simulation. Current from each element with particular X-ray field was calculated. To evaluate the X-ray spectrum expectation from current readout are evaluated with two ways, one is using response matrix solution, and the other is unfolding with SAND-II. It looks represent the iodine effective thickness.
Otaka, Masahiko; Nagai, Keiichi; Ara, Kuniaki
no journal, ,
no abstracts in English
Hashimoto, Makoto; Mikami, Kenta*; Uesaka, Akio*; Kanno, Ikuo; Otaka, Masahiko; Ara, Kuniaki; Nomiya, Seiichiro*; Onabe, Hideaki*
no journal, ,
X-ray energy spectrum measurement in a short time is necessary for CT picture measurement with energy difference method. For this purpose, current mode detector, which is applied X-ray detectors used with current mode and can obtain rough X-ray energy spectrum, is examined. X-ray detector array is used as the electric current mode detector and unfolding of output of the detector to energy spectrum is studied. Furthermore, optimized detector array constitution is also discussed.