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Torikawa, Tomoaki*; Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki
Konsoryu, 36(1), p.63 - 69, 2022/03
On free surface of a sodium cooled fast reactor, gas entrainment can be caused by free surface vortices, which may result in disturbance in core power. It is important to develop an evaluation model to predict accurately entrained gas flow rate. In this study, entrained gas flow rate a simple gas entrainment experiment is conducted with focusing on effect of pressure difference between upper and lower tanks. Pressure difference between upper and lower tanks are controlled by changing gas pressure in lower tank. As a result, it is confirmed that the entrained gas flow rate increases with increasing pressure difference between upper and lower tanks. By visualization of swirling annular flow in suction pipe, it is also observed that pressure drop in suction pipe increases with increase in entrained gas flow rate, which implies that entrained gas flow rate can be predicted by evaluation model based on pressure drop in swirling annular flow region.
Haga, Katsuhiro; Kogawa, Hiroyuki; Naoe, Takashi; Wakui, Takashi; Wakai, Eiichi; Futakawa, Masatoshi
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
The cross-flow type target was developed as the basic design of mercury target in J-PARC, and the design has been improved to realize the MW-class pulsed spallation neutron source. When the high-power and short-pulsed proton beam is injected into the mercury target, pressure waves are generated in mercury by rapid heat generation. The pressure waves induce the cavitation damages on the target vessel. Two countermeasures were adopted, namely, the injection of microbubbles into mercury and the double walled structure at the beam window. The bubble generator was installed in the target vessel to absorb the volume inflation of mercury and mitigate the pressure waves. Also, the double walled target vessel was designed to suppress the cavitation damage by the large velocity gradient of rapid mercury flow in the narrow channel of double wall. Finally, we could attain 1 MW beam operation with the duration time of 36.5 hours in 2020, and achieved the long term stable operation with 740 kW from April in 2021. This report shows the technical development of the high-power mercury target vessel in view of thermal hydraulics to attain 1 MW operation.
Takeda, Takeshi
JAEA-Data/Code 2021-006, 61 Pages, 2021/04
JAEA-Data-Code-2021-006.pdf:2.78MB
An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.
Takeda, Takeshi
JAEA-Data/Code 2018-003, 60 Pages, 2018/03
JAEA-Data-Code-2018-003.pdf:3.68MB
Experiment SB-PV-07 was conducted on June 9, 2005 using LSTF. Experiment simulated 1% pressure vessel top small-break LOCA in PWR under total failure of HPI system and nitrogen gas inflow to primary system from ACC tanks. Liquid level in upper-head was found to control break flow rate. Coolant was started to manually inject from HPI system into cold legs as first accident management (AM) action when maximum core exit temperature reached 623 K. Fuel rod surface temperature largely increased because of late and slow response of core exit temperature. SG secondary-side depressurization was initiated by fully opening relief valves as second AM action when primary pressure decreased to 4 MPa. However, second AM action was not effective on primary depressurization until SG secondary-side pressure decreased to primary pressure. Pressure difference became larger between primary and SG secondary sides after ACC tanks started to discharge nitrogen gas.
4 bundle at 2 MPaLiu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki
Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06
To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 4 bundle under 1.6 MPa (202 
C), 2.1 MPa (215 C) and 2.6 MPa (226 C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.
Kogawa, Hiroyuki; Ishikura, Shuichi*; Sato, Hiroshi; Harada, Masahide; Takatama, Shunichi*; Futakawa, Masatoshi; Haga, Katsuhiro; Hino, Ryutaro; Meigo, Shinichiro; Maekawa, Fujio; et al.
Journal of Nuclear Materials, 343(1-3), p.178 - 183, 2005/08
Times Cited Count：8 Percentile：49.16(Materials Science, Multidisciplinary)A cross-flow type (CFT) mercury target with flow guide blades, which has been developed for JSNS, can suppress the generation of stagnant flow region especially near the beam window where the peak heat density is generated due to spallation reaction. Then, a flat type beam window has been applied to the CFT target from the viewpoint of suppressing dynamic stress caused by a pressure wave, which has been estimated with a mercury model of the linear equation of state. The recent experimental results obtained by using a proton beam incidents to mercury led that a cutoff pressure model in the equation of state of mercury caused a suitable dynamic stress with experimental results. Dynamic stress analyses were carried out with the cutoff pressure model, in which the negative pressure less than 0.15 MPa was not generated. The generated dynamic stress in the flat beam window became much larger than that in a semi-cylindrical type window. However, the generated stress in the semi-cylindrical type beam window was over the allowable stress of SS316L under the peak heat density of 668 W/cc. In order to decrease the dynamic stress in the semi-cylindrical beam window, the incident proton beam was defocused to decrease the peak heat density down to 218 W/cm
. As a result, the dynamic stress could be suppressed less than the allowable stress. On the other hand, due to defocus of the proton beam, high heat density was generated on the end of the flow guide blades, which caused high thermal stress exceeding the allowable stress. To decrease the thermal stress, several shapes of the blade ends were studied analytically, which were selected so as not to affect the mercury flow distribution. A simple thin-end blade showed low thermal stress below the allowable stress.
Fujimoto, Nozomu; Tachibana, Yukio; Saikusa, Akio*; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.273 - 281, 2004/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.
Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi
JAERI-Tech 2004-035, 18 Pages, 2004/03
JAERI-Tech-2004-035.pdf:0.81MB
Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO
blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.
Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00
Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.
Taniguchi, Masaki; Hanada, Masaya; Iga, Takashi*; Inoue, Takashi; Kashiwagi, Mieko; Morishita, Takatoshi; Okumura, Yoshikazu; Shimizu, Takashi; Takayanagi, Tomohiro; Watanabe, Kazuhiro; et al.
Nuclear Fusion, 43(8), p.665 - 669, 2003/08
Times Cited Count：16 Percentile：46.47(Physics, Fluids & Plasmas)no abstracts in English
Taniguchi, Masaki; Hanada, Masaya; Iga, Takashi*; Inoue, Takashi; Kashiwagi, Mieko; Morishita, Takatoshi; Okumura, Yoshikazu; Shimizu, Takashi*; Takayanagi, Tomohiro*; Watanabe, Kazuhiro; et al.
Proceedings of 19th IAEA Fusion Energy Conference (FEC 2002) (CD-ROM), 5 Pages, 2002/10
A high power neutral beam injector (NBI) has been designed for ITER. A key component of the NBI system is a high power beam source which produces a 40A D
ion beams at the energy of 1 MeV. JAERI has developed a vacuum insulated beam source (VIBS). The VIBS insulates the high voltage of 1 MV by immersing the ion source and accelerator in vacuum. So far the VIBS succeeded in acceleration of 37 mA (power supply drain current) beam up to 970 keV for 1 s. The achieved beam energy is nearly equal to the required value for the ITER NBI system. The negative ion source for the ITER beam source has been also developed. One of the key issues for the negative ion source is reduction of the operating pressure. By optimizing the filter magnetic field for negative ion production even at low pressure, a H ion beam of 31 mA/cm
was extracted at 0.1 Pa. Although the pulse length was very short (0.1 s) the ITER requirement on the current density was demonstrated at 1/3 of the ITER design pressure (0.3 Pa), which could reduce the heat loading on the accelerator grids.
Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nagatomi, Hideki; Kaminaga, Masanori; Funayama, Yoshiro
JAERI-Tech 2002-034, 40 Pages, 2002/03
JAERI-Tech-2002-034.pdf:1.97MB
JRR-4, a swimming-pool type research reactor with a thermal power of 3.5MW, attained criticality in July 1998, after replacing its 90% enrichment fuel with a 20% enrichment fuel under the Reduced Enrichment Program. As a part of the program, safety analysis on thermo-hydraulics of the reactor core was conducted on cases including single channel blockage accident. With the conclusion that a certain margin on thermo-hydraulics was necessary, investigation and experiments were carried out with an aim to increase the core flow rate. To increase the core flow, it was carried out to reduce the bypass flow in the core and to increase the primary coolant flow rate from 7m/min to 8m/min. After flow measurements using a mock-up fuel element, flow velocity of the fuel channel was determined as 1.45m/s as opposed to the designed value of 1.44m/s, and the ratio of core flow to total flow was 0.88, exceeding the value 0.86 used for the safety analysis.This report describes the JRR-4 core flow increase plan as well as the results of the channel flow rate measurement
Hamada, Kazuya; Kato, Takashi; Kawano, Katsumi; Hara, Eiji*; Ando, Toshinari; Tsuji, Hiroshi; Okuno, Kiyoshi; Zanino, L.*; Savoldi, L.*
AIP Conference Proceedings 613, p.407 - 414, 2002/00
no abstracts in English
Hanada, Masaya; Kashiwagi, Mieko; Morishita, Takatoshi; Taniguchi, Masaki; Okumura, Yoshikazu; Takayanagi, Tomohiro; Watanabe, Kazuhiro
Fusion Engineering and Design, 56-57, p.505 - 509, 2001/10
Times Cited Count：18 Percentile：76.49(Nuclear Science & Technology)no abstracts in English
Takase, Kazuyuki; Ose, Yasuo*; Akimoto, Hajime
Proceedings of the 1st International Symposium on Advanced Fluid Information (AFI-2001), p.227 - 232, 2001/10
no abstracts in English
Takase, Kazuyuki; Akimoto, Hajime
Applied Electromagnetics in Materials, p.177 - 178, 2001/00
no abstracts in English
Yonomoto, Taisuke; Otsu, Iwao
Proceedings of 12th Pacific Basin Nuclear Conference (PBNC 2000), Vol.1, p.317 - 329, 2000/00
no abstracts in English
Takase, Kazuyuki; Ose, Yasuo*; Akimoto, Hajime
Nihon Kikai Gakkai Dai-12-Kai Keisan Rikigaku Koenkai Koen Rombunshu, p.395 - 396, 1999/00
no abstracts in English
Mori, Yuki*; Kato, Sayuri*; Mori, Hiroko*; Katayama, Yoshinori; Tsuji, Kazuhiko*
Review of High Pressure Science and Technology, 7, p.353 - 355, 1998/03
The temperature dependence of phase transitions in GaSb, AlSb, GaAs, GaP, InAs, ZnSe, and CdTe are studied by X-ray diffraction measurements under pressure upto 30 GPa at temperatures of 90-300K. The phase transitions depend on paths in a pressure-temperature phase diagram. The structure of the recovered phase after decompression depends on the ionicity in bonding: amorphous for small ionicity, the stable zincblende structure for large ionicity, and microcrystalline or moderate ionicity. These results are discussed by using a configuration-coordinate model.
Toda, Naohiro*; Katayama, Yoshinori; Tsuji, Kazuhiko*
Review of High Pressure Science and Technology, 7, p.647 - 649, 1998/03
The electrical conductivity 
has been measured at pressures 
to 8 GPa and temperatures 
of 77-300K in evaporated amorphous Ge (a-Ge), a-Ge-Cu alloys and a-Ge-Al alloys. The dependence of is well described by a power lw at low temperatures below 150 K, which is expemcted from a multi-phonon tunneling transition process model with weak electron-lattice coupling, rather than the Mott's variable range hopping conduction model. The exponent 
in the power law changes with increasing pressure. For both a-Ge
Cu
and a-GeAl alloys, d(ln )/d show positive values in the low pressure region and negative values in the high pressure region. Results are discussed from several hopping conduction models.
