Kuroda, Kenta*; Arai, Yosuke*; Rezaei, N.*; Kunisada, So*; Sakuragi, Shunsuke*; Alaei, M.*; Kinoshita, Yuto*; Bareille, C.*; Noguchi, Ryo*; Nakayama, Mitsuhiro*; et al.
Nature Communications (Internet), 11, p.2888_1 - 2888_9, 2020/06
Hayakawa, Sho*; Doihara, Kohei*; Okita, Taira*; Itakura, Mitsuhiro; Aichi, Masaatsu*; Suzuki, Katsuyuki*
Journal of Materials Science, 54(17), p.11509 - 11525, 2019/09
Hayakawa, Sho*; Okita, Taira*; Itakura, Mitsuhiro; Kawabata, Tomoya*; Suzuki, Katsuyuki*
Journal of Materials Science, 54(16), p.11096 - 11110, 2019/08
Nakanishi, Daiki*; Kawabata, Tomoya*; Doihara, Kohei*; Okita, Taira*; Itakura, Mitsuhiro; Suzuki, Katsuyuki*
Philosophical Magazine, 98(33), p.3034 - 3047, 2018/09
By using the six sets of interatomic potentials for face-centredcubic metals that differ in the stacking fault energy (SFE) while most of the other material parameters are kept almost identical, we conducted molecular dynamics simulations to evaluate the effects of SFE on the defect formation process through collision cascades. The ratio of glissile SIA clusters tends to decrease with increasing SFE. This is because perfect loops, the edges of which split into two partial dislocations with stacking fault structures between them in most cases, prefer to form at lower SFEs. The enhanced formation of glissile SIA clusters at lower SFEs can also be observed even at increased temperature.
Hayakawa, Sho*; Okita, Taira*; Itakura, Mitsuhiro; Aichi, Masaatsu*; Suzuki, Katsuyuki*
Philosophical Magazine, 98(25), p.2311 - 2325, 2018/06
We conduct kinetic Monte Carlo simulations for the conservative climb motion of a cluster of self-interstitial atoms towards another SIA cluster in BCC Fe; the conservative climb velocity is inversely proportional to the fourth power of the distance between them, as per the prediction based on Einstein's equation. The size of the climbing cluster significantly affects its conservative climb velocity, while the size of the cluster that originates the stress field does not. The activation energy for the conservative climb is considerably greater than that derived in previous studies and strongly dependent on the climbing cluster size.
Doihara, Kohei*; Okita, Taira*; Itakura, Mitsuhiro; Aichi, Masaatsu*; Suzuki, Katsuyuki*
Philosophical Magazine, 98(22), p.2061 - 2076, 2018/05
In this study, molecular dynamics simulations were performed to elucidate the effects of stacking fault energy (SFE) on the physical interactions between an edge dislocation and a spherical void in the crystal structure of face-centred cubic metals at various temperatures and for different void sizes. Four different types of interaction morphologies were observed, in which (1) two partial dislocations detached from the void separately, and the maximum stress corresponded to the detachment of the trailing partial; (2) two partial dislocations detached from the void separately, and the maximum stress corresponded to the detachment of the leading partial; (3) the partial dislocations detached from the void almost simultaneously without jog formation; and (4) the partial dislocations detached from the void almost simultaneously with jog formation. With an increase in void size or SFE, the interaction morphology changed in the above-mentioned order. It was observed that the magnitude of the critical resolved shear stress (CRSS) and its dependence on the SFE were determined by these interaction morphologies. The value of the CRSS in the case of interaction morphology (1) is almost equal to an analytical one based on the linear elasticity by employing the Burgers vector of a single partial dislocation. The maximum value of the CRSS is also obtained by the analytical model with the Burgers vector of the two partial dislocations.
Kuroda, Kenta*; Ochi, Masayuki*; Suzuki, Hiroyuki*; Hirayama, Motoaki*; Nakayama, Mitsuhiro*; Noguchi, Ryo*; Bareille, C.*; Akebi, Shuntaro*; Kunisada, So*; Muro, Takayuki*; et al.
Physical Review Letters, 120(8), p.086402_1 - 086402_6, 2018/02
Ito, Daisuke*; Rivera, M. N.*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 10 Pages, 2017/09
Nava, M.*; Ito, Daisuke*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 5 Pages, 2017/07
Ito, Daisuke*; Nava, M.*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 2017 Japan-US Seminar on Two-Phase Flow Dynamics (JUS 2017), 4 Pages, 2017/06
Aoyagi, Mitsuhiro; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 14 Pages, 2014/12
A numerical model for freezing and blockage formation of solid-liquid flow in the SIMMER code was validated in order to improve the accuracy in evaluating fuel discharge behavior in the core disruptive accident of FBR. The THEFIS experiment which investigated fuel discharge behavior was chosen as reference data in this study. The numerical conditions were set according to the experimental system. Although the experimental result was well simulated by using the existing numerical model of SIMMER, the melt flow was suppressed excessively in some cases. Overestimation of flow resistance by the solid particles in the numerical model was discovered though the comparison between the numerical model and the physical phenomenon in the experiment. The numerical model caused the excessive melt flow suppression. Therefore, we improved the numerical model to adapt to the actual phenomenon. Then, it was confirmed the improved numerical model brought more appropriate numerical results.
Suzuki, Mitsuhiro; Nakamura, Hideo
Journal of Nuclear Science and Technology, 47(12), p.1193 - 1205, 2010/12
Presented in the paper are experimental results on general performance of core exit thermocouple (CET) to detect core overheat for accident management (AM) action. Thirteen tests simulating small break loss-of-coolant accident (SBLOCA) and abnormal transient are studied by using the Large Scale Test Facility (LSTF) which is a full-height, full-pressure and 1/48 volumetric-scaled PWR model. Clarified are as follows, (1) general CET performance with certain delay in time and temperature rise from core overheating in most cases, (2) one common reason of the delay due to cooling effects of metal structures in core and core exit, (3) an indication of superheat instead of its temperature necessary for significantly high or low pressure transients, (4) no CET heat-up in case of large water fall-back from hot legs and in addition, discussion on applicability to PWR is presented.
Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo
JAEA-Research 2009-057, 188 Pages, 2010/02
A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility of ROSA-V Program to have an insight into effects of accident management action on core cooling during a simulated vessel top break loss-of-coolant accident with a total failure assumption on the high pressure injection (HPI) system at a pressurized water reactor (PWR). Typical phenomena of vessel top break with break sizes between 1.0 and 0.1% cold leg break equivalent were clarified including upper head water level transients related to steam discharge, coolant mass inventory related to core heat-up, performance of core exit thermocouple (CET)and three-dimensional steam flows in core and core exit. Both operator actions of HPI recovery in the 1.0% top break and steam generator depressurization in the 0.1% top break resulted in immediate recovery of core cooling when these were initiated by CET heat-up at 623 K.
Suzuki, Mitsuhiro; Nakamura, Hideo
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 17 Pages, 2009/09
Presented are experiment results on performance of core exit thermocouple (CET) and applicability to PWR accident management (AM) during 12 tests of small-break loss-of-coolant accident (SBLOCA) and abnormal transient conducted at the Large Scale Test Facility (LSTF) of Japan Atomic Energy Agency, which is the largest PWR simulator with full-height and 1/48 volume scaling. General CET performances are derived including (1) CETs are capable in most cases to detect core overheating with delay of time and temperature increase from core heat-up, (2) one of the reasons of this delay is attributed to cooling effects of structural materials at the core exit and peripheral region, (3) CETs were incapable to detect core overheating in a very small break under steam generator depressurization action as well as a 10% cold leg break due to significant water fall-back from hot legs, and (4) an alternative indication by CET superheat is necessary in extremely high and low pressure conditions.
Suzuki, Mitsuhiro; Nakamura, Hideo
JAEA-Research 2009-011, 155 Pages, 2009/07
This report summarizes performances of core exit thermocouples (CETs) observed in 12 ROSA/LSTF tests which include ten small-break loss-of-coolant accident (SBLOCA) tests and two abnormal transient tests as an additional report to the OECD/NEA ROSA Project Test 6-1 report. The contents of this report are prepared to a task group which was set up to review and consolidate background knowledge of CET application to PWR accident management (AM) measures in April 2008 in the Working Group of Analysis and Management of Accident (WGAMA) at OECD/NEA. These tests cover wide ranges of test conditions such as size and location of break, primary pressure, core power, reflux water fall-back and operator actions. CET performances relative to the core temperature history are studied in each test, and general performances of CET are summarized focusing on the time delay and slow and low temperature excursion.
Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 3(1), p.146 - 157, 2009/00
Presented are experiment results of the LSTF with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break LOCA simulation experiment. The break size is equivalent to 1.9% cold leg break. The accident management (AM) action to rapidly open the SG relief valves was initiated when CET temperature rose up to 623 K. The core overheat, however, was detected with a time delay of about 230 s and a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarified the reasons of time delay and temperature discrepancy between the CETs and heated core including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to PWR conditions and a possibility of alternative indicators for earlier AM action is studied by using symptom-based plant parameters such as a reactor vessel water level detection.
Suzuki, Mitsuhiro; Nakamura, Hideo
JAEA-Research 2008-087, 148 Pages, 2008/10
This report presents major results observed in LOCA test (SB-CL-09) conducted at the ROSA/LSTF test facility simulating 10% cold leg break in a 4-loop Westinghouse-type PWR. Following are found in this test with an assumption of high pressure injection system. (1) The relatively large break size resulted in pressure inverse within 2 minutes between the primary and steam generator secondary sides. (2) During a loop-seal clearing (LSC) process started at about 1 minutes after the break, the core water level was suppressed to almost lower end and then it recovered to the middle core height. The water level remained at the middle height was due to remained water levels in the SG U-tube inlet sides which were higher than their outlet sides. (3) Significant core heat-up was observed above the water level at the middle height and core power was tripped off at 111s. (4) The effects of fall-back water from the intact loop hot leg was observed by the local core cooling.
Iba, Katsuyuki*; Ozeki, Takahisa; Totsuka, Toshiyuki; Suzuki, Yoshio; Oshima, Takayuki; Sakata, Shinya; Sato, Minoru; Suzuki, Mitsuhiro; Hamamatsu, Kiyotaka; Kiyono, Kimihiro
Fusion Engineering and Design, 83(2-3), p.495 - 497, 2008/04
Fusion research grid is an environment of collaborative researches using a network that connects scientists far apart and let them collaborate effectively over the difference in time and distance in a nuclear fusion research. Fundamental technology of Fusion research grid has been developed at JAEA in the VizGrid project under the e-Japan project at Ministry of Education, Culture, Sports, Science and Technology (MEXT). Remote research environments of experiments, diagnostics, analyses and communications were developed on Fusion research grid. We have developed prototype systems that include a remote experiment system, a remote diagnostics system, and a remote analysis system. All users can access these systems from anywhere because Fusion research grid does not required closed network like Super SINet to maintain security. The prototype systems were verified in experiments at JT-60U and their availability was confirmed.
Takeda, Takeshi; Asaka, Hideaki; Suzuki, Mitsuhiro; Nakamura, Hideo
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 3 Pages, 2007/09
RELAP5 code analysis was performed to validate the code predictability by using ROSA/LSTF experiment data that simulated a PWR vessel upper head small break loss-of-coolant accident (SBLOCA) with a break equivalent to 1% cold leg break. The JAEA-modified RELAP5/MOD18.104.22.168 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient (Cd) of 0.61 for two-phase break flow. In the experiment, liquid level in the upper head was found to control break flow rate as coolant in the upper plenum entered the upper head through control rod guide tubes (CRGTs) until the penetration holes at the CRGT bottom were exposed to steam in the upper plenum. The upper head noding and flow paths between the upper plenum and the CRGT were thus modeled to simulate well the liquid level and coolant flow around the upper portion of pressure vessel. The code, however, overpredicted the break flow rate due to the underprediction of break-upstream void fraction especially during two-phase flow discharge period. Cd for two-phase break flow was thus adjusted to be 0.58. Effects of break area on the core cooling were investigated further. The parameter analyses showed that peak cladding temperature (PCT) is the maximum at 1% break case, while the PCT would be lower than 1200 K in the larger break size cases because vapor condensation on injected accumulator coolant induces loop seal clearing and effectively enhances core cooling thereafter.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
JAEA-Research 2007-037, 150 Pages, 2007/03
A small break LOCA experiment (SB-PV-06) was conducted at the LSTF of ROSA-V program to study effects of rapid secondary depressuriza-tion action on core cooling as one of accident management (AM) measures for a PWR in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. The break simulated 10 instrument tubes rupture equivalent to 0.2% cold leg break. It was clarified through comparison with former experiments that (1) the depressurization initiated by detecting the vessel level below the primary loop (4545s) was degraded by the gas inflow resulting in whole core uncovery prior to the start of low pressure injection and (2) an alternative start of the depressurization by detecting level decrease at the SG outlet plenum (2330s), would limit the core uncovery suggesting more effective parameter for the AM measures. The report presents the experiment results with the effects of rapid depressurization initiation timing.