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Journal Articles

Internal strain distribution of laser lap joints in steel under loading studied by high-energy synchrotron radiation X-rays

Shobu, Takahisa; Shiro, Ayumi*; Kono, Fumiaki*; Muramatsu, Toshiharu; Yamada, Tomonori; Naganuma, Masayuki; Ozawa, Takayuki

Quantum Beam Science (Internet), 5(2), p.17_1 - 17_9, 2021/06

The automotive industries employ laser beam welding because it realizes a high energy density without generating irradiation marks on the opposite side of the irradiated surface. Typical measurement techniques such as strain gauges and tube X-rays cannot assess the localized strain at a joint weld. Herein high-energy synchrotron radiation X-ray diffraction was used to study the internal strain distribution of laser lap joint PNC-FMS steels (2- and 5-mm thick) under loading at a high temperature. As the tensile load increased, the local tensile and compressive strains increased near the interface. These changes agreed well with the finite element analysis results. However, it is essential to complementarily utilize internal defect observations by X-ray transmission imaging because the results depend on the defects generated by laser processing.

Journal Articles

Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor

Imaizumi, Yuya; Yamada, Fumiaki; Arikawa, Mitsuhiro*; Yada, Hiroki; Fukano, Yoshitaka

Mechanical Engineering Journal (Internet), 5(4), p.18-00083_1 - 18-00083_11, 2018/08

A calculation program was developed to evaluate and discuss the effectiveness of the countermeasures such as sodium pump-up and siphon-breaking against the loss-of-reactor-level (LORL) where the coolant circulation path is lost in loop-type sodium-cooled fast reactors. Due to the non-negligible possibility obtained by probabilistic risk assessment (PRA), sodium leakages in two points both occurred in primary heat transport system (PHTS) was assumed in this study. In addition, the crack size was discussed and evaluated realistically, instead of the value that was assumed in the conventional studies. Representative sequences and leakage positions were chosen, and the sodium level transient in reactor vessel (RV) was calculated. The calculations were also conducted where the larger crack size was set for the second leakage, in order to investigate additional requirements to maintain the RV sodium level. The evaluation results clarified that the coolant circulation loop can be maintained even after the second leakage in PHTS, taking into account the effects by the countermeasures.

Journal Articles

Development of the severe accident evaluation method on second coolant leakages from the PHTS in a loop-type sodium-cooled fast reactor

Yamada, Fumiaki; Imaizumi, Yuya; Nishimura, Masahiro; Fukano, Yoshitaka; Arikawa, Mitsuhiro*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07

The loss-of-reactor-level (LORL) is one of the loss-of-heat-removal-system (LOHRS) of beyond-DBA (BDBA) severe accident. An evaluation method for the LORL which is caused by the coolant leakage in two positions of the primary heat transport system (PHTS) was developed for prototype JSFR which is loop-type sodium-cooled fast reactor. The secondary leakage in cold standby which occurred in different loop from that of the first leakage in rated power operation can lead LORL by excessive declining of the sodium level. Therefore, the sodium level behavior in RV was studied in a representative accident sequence by considering the sodium pumping up into RV, siphon-breaking to stop pumping out from RV and maintain the sodium level, and calculation programs for the transient sodium level in RV. The representative sequence with lowest sodium level was selected by considering combinations of possible leakage positions. As a result of the evaluation considering the countermeasures above, it was revealed that the LOHRS can be prevented by maintaining the sodium level for the operation of decay heat removal system, even in the leakages in two positions of PHTS which corresponds to BDBA.

Journal Articles

Safety margins after failure of fuel cladding during protected loss-of-heat-sink accidents in a sodium-cooled fast reactor

Fukano, Yoshitaka; Nishimura, Masahiro; Yamada, Fumiaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5687 - 5698, 2015/08

The following safety criteria for anticipated operational occurrences are commonly and uniformly employed for all the DBAs in the Japanese prototype sodium-cooled fast reactor to prevent fuel melting and cladding failure:(a) Maximum fuel temperature shall be below the melting point,(b) Maximum cladding temperature shall be below 830$$^{circ}$$C, and (c) Maximum coolant temperature shall be below the boiling point. Cladding failure is allowed, on the contrary to that, in beyond DBAs (BDBAs) or severe accidents (SAs), whereas the core cooling capability is also needed to be secured as in DBAs. No fuel melting enables this by keeping the core in a coolable geometry, and is thus conservatively required even under such a condition. Protected loss-of-heat-sink (PLOHS) events are identified as one of the most dominant sequences. Safety margins for significant core damage in PLOHS events were therefore studied in this paper assuming fuel cladding failure. The following three possible mechanisms leading to degradation of the core were then identified to be scrutinized by a thorough and state-of-the-art review of open papers on the phenomena anticipated to occur under cladding failure conditions:(1) Fuel melting due to fuel-sodium reaction product (FSRP) formation, (2) Thermal transient due to FP gas impingement from adjacent failed fuel pins, and (3) Mechanical load due to the same FP gas impingement. It was clarified through simulation analyses on each phenomenon mentioned above using the FUCA code that there was no significant core damage at the coolant temperatures of up to 950$$^{circ}$$C. It was therefore concluded that large safety margins are provided during PLOHS events even in failure of fuel cladding.

Journal Articles

Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

JAEA Reports

A Study on laser welding of ferritic/martensitic steel (PNC-FMS) for fast reactor fuel assemblies

Kono, Fumiaki; Sogame, Motomu; Yamada, Tomonori; Shobu, Takahisa; Naganuma, Masayuki; Ozawa, Takayuki; Muramatsu, Toshiharu

JAEA-Technology 2015-004, 57 Pages, 2015/03


Laser welding of ferritic/martensitic steel (PNC-FMS) sheets with different thicknesses (2 mm and 5 mm) was examined to investigate the weldability between the inner and outer duct in fast reactor fuel assemblies with inner duct structure (FAIDUS); the objective of the inner duct is to avoid the re-criticality in case of the core melting accident. Laser-spot and melt-run welding was performed at various laser powers, welding times and velocities to find out the appropriate welding conditions with few defects and enough penetration depth. As for the spot welding, furthermore, slow cooling rate or pulsed laser irradiation could reduce the crack and porosity in the welded zone. The strain of the welded zone almost disappeared and the hardness was comparable with that of the base metal by applying post welding heat treatment at 690 $$^{circ}$$C for 103 min. In addition, the shear strength of welded joints was confirmed to be sufficiently higher than the provisional allowance shear stress. These results indicate that laser welding would be probably applied to the PNC-FMS inner and outer ducts.

Journal Articles

Development of natural circulation analytical model in Super-COPD code and evaluation of core cooling capability in Monju during a station blackout

Yamada, Fumiaki; Fukano, Yoshitaka; Nishi, Hiroshi; Konomura, Mamoru

Nuclear Technology, 188(3), p.292 - 321, 2014/12

 Times Cited Count:13 Percentile:77.66(Nuclear Science & Technology)

The capability of natural circulation for core cooling has been evaluated in detail for a station blackout (SBO) event induced by an earthquake and a subsequent tsunami hit. The evaluation was prompted by the accident at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company. The plant dynamics computer code Super-COPD was used for the evaluation, which has been validated by analyses of preliminary test results on the natural circulation in Monju. As a result, it was concluded that natural circulation of the sodium coolant will enable the decay heat from the core to be removed under such an SBO condition.

Journal Articles

Evaluation on coolability of the reactor core in Monju by natural circulation under earthquake and subsequent tsunami event

Yamada, Fumiaki; Fukano, Yoshitaka; Nishi, Hiroshi; Konomura, Mamoru

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

The core cooling capability by natural circulations at a station black-out event, induced by an earthquake and a subsequent tsunami attack, has been evaluated in detail, referring to the accident of the Fukushima Dai-ichi Nuclear Power Station of Tokyo Electric Power Company. The plant dynamics computer code: Super-COPD has been used for the evaluation, which has been verified by the analyses of the preliminary test results on the natural circulation in Monju. As a result it was concluded that the natural circulations of the coolant sodium will enable the decay heat removal of the core as far as the sodium coolant flow circuits are intact and secured.

JAEA Reports

Development of the Monju core safety analysis numerical models by Super-COPD code

Yamada, Fumiaki; Minami, Masaki*

JAEA-Data/Code 2010-023, 79 Pages, 2010/12


Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model.

Journal Articles

Numerical simulation of MONJU plant dynamics by Super-COPD using previous startup tests data

Yamada, Fumiaki; Ohira, Hiroaki

Proceedings of 3rd Joint US-European Fluids Engineering Summer Meeting and 8th International Conference on Nanochannels, Microchannels, and Minichannels (ASME 2010) (CD-ROM), 8 Pages, 2010/08

The advanced flow network models of the RV upper plenum and the IHX inlet plenum of MONJU were explained and validated by the previous SSTs. The whole plant dynamics of MONJU were also predicted using the validated flow networks. The natural circulation experiments both in the PHTS and the SHTS were conducted applying the previous SST conditions. The whole plant dynamics model with the advanced IHX model was also validated by these test results. Through these validations, we concluded that the present plant dynamics model of Super-COPD could simulate the whole plant dynamics in good accuracy, which was applicable to the next SSTs.

JAEA Reports

Development of grout materials for a geological disposal system for high-level radioactive waste

Fukuoka, Naomi; Shinkai, Fumiaki; Miura, Norihiko*; Nobuto, Jun*; Yamada, Tsutomu*; Naito, Morimasa

JAEA-Data/Code 2010-005, 353 Pages, 2010/07


High-level radioactive waste management in Japan is based on the multi-barrier concept, composed of the engineered barrier system and the surrounding geological formations. Although cementitious materials are commonly used for rock support, lining, and grouting, their pH plume are considered to have an adverse effect on long-term safety of a geological disposal system. In addition, during the emplacement of waste package with buffer material, it is required to limit amount of groundwater inflow into a disposal pit or tunnel to a certain level by grouting because the bentonite clay buffer is easy to swell in time by contact with the groundwater. Therefore, it is necessary to develop new grout materials with penetrability for smaller fractures. This report shows the most appropriate composition of new grout materials to be suitable for the in-situ experiment based on the result of indoor test.

Journal Articles

Unplanned shutdown frequency prediction of FBR MONJU using fault tree analysis method

Sotsu, Masutake; Yamada, Fumiaki*

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05

MONJU is a sodium cooled, loop-type prototype fast breeder reactor which can supply 280MW of electricity to the grid. The generated heat at the reactor core is removed by three loops of primary heat transport system (PHTS), each of those is thermally connected through individual intermediate heat exchanger (IHX) to another clloant eirculation loop of secondary heat transport system (SHTS). The turbine generator is driven by steam generated at three evaporators and super heaters installed at the SHTS.

JAEA Reports

Development of "FBR plant engineering system"

Minami, Masaki*; Sakata, Hideaki; Yoshikawa, Shinji; Yamada, Fumiaki

JNC TN4410 2005-001, 123 Pages, 2005/03


A software system for straightforward and quick conceptual studies and technical evaluations of fast breeder reactor plants has been developed, mainly targeting the Japanese Demonstration Fast Breeder Reactor Monju. The studies and evaluations by this system used to be limited within steady and nominal conditions, excluding influences by changing specification values in accidental conditions. In this fiscal year, a new software component has been included in the system for simplified evaluation of transient characteristics, which is an essential for complete design of an FBR plant. This new evaluation function enabled to detect specifications to vary over acceptable range in transient conditions, and to notify necessity for re-adjustment of steady state design specification values. This system was also utilized to generate a sensitivity survey program in order to evaluate appropriateness of Monju design. The appropriateness of Monju design was evaluated in two ways. The first one is to follow specification selecting sequence as the as-built Monju has actually been designed. The second one is hypothetical deviations of major specifications of Monju and observation of the influences on other specifications. As a result, Monju design was confirmed to be adequate from the view point of need to meet design limitations at the design stage of Monju. This system is expected to be systematically re-arranged, because the system has now considerably complicated configuration after additions of various programs for many objects. This arrangement will facilitate future contribution of this system to technical studies in order for development of FBRs.

Journal Articles

Realistic Safety Margin Analysis of Monju Based on Plant Performance Measurements

Yamada, Fumiaki; Kitamura, Kenji*

Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 0 Pages, 2004/04


JAEA Reports

Development of inspection technique for future commercialized FBRs

; Yamada, Fumiaki

JNC TN4400 2003-001, 81 Pages, 2003/06


To achieve the availability factor of the Light Water Reactors (LWRs) for the commercialized Fast Breeder reactors (FBRs) in the future, it is necessary to shorten the period inspection. For this purpose, it is important to develop to develop and to verify inspection techniques for the FBR in the state of filling with opaque liquid sodium. From this point of view, we have investigated the volumetric inspection technique for steam generator (SG) tube at high temperature and the under-sodium visual inspection technique as a part of feasibility study for future commercialized EBRs. The results are as follows: (1)Development of under-sodium Visual Inspection Technique. In order to verify the deterioration of the special resolution, the ability of visualization of ultrasonic wave sensor in an ideal environmental condition as in a reactor vessel have been evaluated by analytic method. As a result, the effects from the signal to noise (S/N) ratio of sensor elements and the angle between the sensor and the target are smaller, and that from the temperature fluctuation in the reactor vessel is bigger. Though ASME section IX rule requires to identify the English lower-case letters with size 2.7mm at the distance of 1.2m, the ultrasound technique used here is hardly to clearly read out the letters at frequency of 10MHz. Therefore, it is necessary to improve the spatial resolution. (2)Development of Volumetric Inspection Technique for Steam Generator Tubes at High Temperature. In order to verify the detectability for flaws, the sensor probes of bobbin ECT and remote field (R/F)-ECT which are composed of heat resistant material and parts have been fabricated. The flaws inspection for 12Cr steel tubes and 2$$cdot$$1/4Cr-1Mo steel tubes and the continuous heat-test for 12Cr steel tubes have been made. The S/N ratios are more than 8 from room temperature to 200$$^{circ}$$C, their detectability has been verified. The eontinuous heat resistance test for more than 100 hours at the temper

Journal Articles

Effect of $$gamma$$-irradiation on microflora and fermentation quality of silages containing tofu cake or brewer's gain

Tanaka, Osamu*; Akiyama, Fumiaki*; Yamada, Akihisa*; Ando, Sada*; Uegaki, Ryuichi*; Kobayashi, Ryoei*; Kume, Tamikazu

Nihon Sochi Gakkai-Shi, 47(3), p.274 - 282, 2001/08

no abstracts in English

Journal Articles

Effect of $$gamma$$ irradiation on microflora and fermentation quality of alfalfa, orchardgrass, maize and sorgum silage

Tanaka, Osamu*; Akiyama, Fumiaki*; Yamada, Akihisa*; Ando, Sada*; Uegaki, Ryuichi*; Kobayashi, Ryoei*; Kume, Tamikazu

Nihon Sochi Gakkai-Shi, 47(1), p.62 - 67, 2001/04

no abstracts in English

Journal Articles

Monju safety margin analysis based on measurements

Nishida, Kazuhiro; Kitamura, Kenji*; Yamada, Fumiaki

Saikuru Kiko Giho, (10), p.5 - 13, 2001/03

The results of the function test and the startup test of the Prototype Fast Breeder Reactor Monju were examined for the hypothetical analysis of the rapid withdrawal of the control rod assembly and to analyze pump seizure in the Primary Heat Transport System, in order to obtain accurate transition of the plant dynamics. These analyses confirmed that both of the accidents settle down without affecting the plant structure. During the withdrawal accident, reactor power reached 106%. During the pump seizure accident, the fuel cladding mid-wall temperature reached 702$$^{circ}$$C. Both of these values are sufficiently below the safety margins for anomaly transients and accidents during operation and former analyses for design safety permission. The differences between these results and the former analyses for design safety permission have mostly been attributed to the characteristics of the control rods.

Journal Articles


Passala; Yamada, Fumiaki

Proceedings of 10th International Conference on Nuclear Engineering (ICONE-10), , 


Oral presentation

Handbook for supposed trouble cases and restore actions in Monju

Yamada, Fumiaki; Tabata, Hiroaki; Miyakawa, Akira; Okada, Mamoru; Okabe, Ayao; Misawa, Naoto*

no journal, , 

no abstracts in English

47 (Records 1-20 displayed on this page)