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Yoshimura, Kazuo; Doda, Norihiro; Nakamine, Yoshiaki*; Fujisaki, Tatsuya*; Igawa, Kenichi*; Iida, Masaki*; Tanaka, Masaaki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
In Japan Atomic Energy Agency, a virtual plant model of the sodium-cooled fast reactor plant composed in a computer is being developed to reduce the development cost, by replacing the experiments to the numerical simulations with coupled analyses of the physical phenomena accounting for the interaction between components under various plant conditions. Through the numerical analysis of the ULOHS test in the U.S. experimental fast reactor named EBR-II, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
The benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-301 and BOP-302R) have been conducted in order to validate the evaluation method of the reactivity feedback equipped in the plant dynamics analysis code named Super-COPD. In this study, 1D-CFD coupled analyses adding the core bowing reactivity model were conducted. Through the analysis, the applicability of the modified reactivity model was confirmed for the BOP-301 test. For the BOP-302R test, consideration of the core restraint system in the core and modeling the control rod driveline expansion reactivity was indicated.
Takubo, Yusaku*; Takayama, Yusuke; Idiart, A.*; Tanaka, Tatsuya*; Ishida, Keisuke*; Fujisaki, Kiyoshi*
Proceedings of 2022 International High Level Radioactive Waste Management Conference (IHLRWM 2022) (Internet), p.906 - 915, 2022/11
no abstracts in English
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/08
Development of evaluation method for cover gas entrainment (GE) by vortices generated at free surface in upper plenum of sodium-cooled fast reactor (SFR) is required. GE evaluation tool, named StreamViewer, based on method using numerical results of three-dimensional computational fluid dynamics analysis for loop-type SFRs has been developed. In this study, modification of evaluation method of StreamViewer to rationalize conservativeness in evaluation results was examined by identifying vortex center lines and calculating three-dimensional distribution of pressure decrease along vortex center lines. The applicability of modified method was checked using water experimental result in rectangular open channel where unsteady vortices are generated. As the result, it was indicated that evaluation results on gas core depth which were excessive in current method were improved in modified method, and it is confirmed that modified method may discriminate onset of GE with appropriate criteria.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08
To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.
Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Fujisaki, Tatsuya*; Murakami, Satoshi*; Vilim, R. B.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
In Japan Atomic Energy Agency, the multilevel simulation system which enables consistent evaluation from the whole plant behavior to the local phenomena is being developed to optimize plant design and enhance the safety of sodium-cooled fast reactors. To validate the coupling method in the MLS system, the 1D-CFD coupling method using Super-COPD for 1D plant dynamics analysis and Fluent for multi-dimensional CFD analysis was applied to the analyses of loss of flow tests in EBR-II. It was confirmed that it could predict multi-dimensional thermal-hydraulic phenomena such as thermal stratification in the upper plenum, Z-shaped pipe, and cold pool, holding the whole plant behavior simultaneously. Moreover, the applicability of the 1D-CFD coupling method to the evaluation of the phenomena in natural circulation conditions was confirmed by comparing the results of the 1D-CFD couple analyses and the measured data.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2021-Nendo Koen Rombunshu (Internet), 4 Pages, 2021/08
For evaluation of gas entrainment phenomenon at free surface in reactor vessel of sodium-cooled fast reactor, the gas entrainment evaluation tool named "Stream Viewer" has been developed. In Stream Viewer, depth of surface vortex dimple is predicted by calculating pressure decrease at the vortex center using velocity distribution around the vortex and Burgers vortex model. In this report, a method to identify continuous vortex center lines from a velocity distribution is newly developed. It becomes possible to evaluate three-dimensional distribution of pressure decrease along vortex center line. Then, the method is validated by applying Stream Viewer to an open channel experiment. As the result, it was confirmed that vortex center lines were successfully identified by the improved Stream Viewer. Moreover, it was also shown that the evaluation accuracy of gas entrainment was expected to be improved by considering distribution of pressure decrease along vortex center line.
Matsushita, Kentaro; Fujisaki, Tatsuya*; Ezure, Toshiki; Tanaka, Masaaki; Uchida, Mao*; Sakai, Takaaki*
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 26, 6 Pages, 2021/05
For the gas entrainment vortex at the free surface in sodium-cooled fast reactors, development of the numerical analysis method to evaluate amount of the gas entrainment from the free surface has been developing. In this paper, the automatic creation of analysis meshes which can suppress the calculation cost while maintaining the prediction accuracy of the vortex shape is investigated, and the adaptive mesh refinement (AMR) method is examined to the creation of analysis mesh applying to the unsteady vortex system. The refined mesh based on the criterion evaluated by vorticity, Q-value as second invariant of the velocity and the discriminant for the eigen equation of the velocity gradient tensor is considered, and it found that the AMR method based on Q-value can refine the analysis meshes most efficiently.
Takaya, Shigeru; Fujisaki, Tatsuya*
Mechanical Engineering Journal (Internet), 7(3), p.19-00526_1 - 19-00526_10, 2020/06
In severe seismic conditions, sloshing waves are considered to even reach a roof slab of a reactor vessel. The structural integrity of roof slabs is required to be evaluated against sloshing impacts. However, there is no widely recognized evaluation method for sloshing impact pressure on flat roofs yet. Therefore, in this paper, a simplified evaluation method is proposed based on Wagner's theory, which is a well-known classic theory for evaluating impact pressures on rigid wedges dropping on water surfaces. In the proposed method, we assume an equivalent wedge on a flat roof. The impact pressure on the equivalent wedge is evaluated by applying Wagner's theory. Computational fluid dynamics analysis is conducted to confirm that a key assumption of Wagner's theory is applicable to the evaluation of sloshing impact on a flat roof. In addition, the predictability of the proposed method is investigated by comparing literature data of sloshing experiments with the estimated values.
Takaya, Shigeru; Fujisaki, Tatsuya*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
Sloshing is one of important issues for both loop-type and tank-type fast reactors with free liquid surface. Periods of seismic vibration are lengthened by base isolation systems installed to prevent damages to facilities during earthquakes, and get close to the natural periods of sloshing. As a result, sloshing is promoted. Sloshing waves are assumed to even reach a roof slab of a reactor vessel in severe seismic conditions. It is important to evaluate structural integrity for sloshing impacts on roofs. However, there are not any established evaluation methods for impact pressure on flat roofs yet. Therefore, in this study, a simple evaluation method is proposed based on Wagner's theory. The effectiveness of the proposed method is illustrated using computational fluid dynamics analysis and literature data of sloshing experiments.
Takaya, Shigeru; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07
Flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design of an advanced loop-type sodium cooled fast reactor. We have been developing numerical analysis models to deal with this issue. In this study, computational fluid dynamics (CFD) simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by the CFD simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.
Takaya, Shigeru; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 5 Pages, 2016/07
JAEA is conducting R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than Monju in order to reduce construction costs. The design increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration is one of main concerns. The flow field in the hot-leg piping is affected by flow disturbance at the inlet, so it is important to evaluate flow field including the upper plenum. In this study, we analyzed unsteady fluid flow by using an integrated model of the upper plenum and the hot-leg piping system. Unsteady Reynolds averaged Navier-Stokes simulation with Reynolds stress model was used. In general, the simulation results obtained by using the integrated model show a similar tendency with the experiment results of 1/3 scaled-model of hot-leg piping with deflect flows. The coupling effect of swirling and deflected flows seems to be not significant although further investigation is needed.
Takaya, Shigeru; Tanaka, Masaaki; Fujisaki, Tatsuya*
Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 7 Pages, 2015/07
Flow-induced vibration of hot-leg pipings is one of concerns for the design of Japan Sodium-cooled Fast Reactor (JSFR) which is now being developed. The flow field in the hot-leg pipings is supposed to be affected by flow disturbances at the entrance, so it is important to evaluate flow fields including the upper plenum. In this study, a simulation model of the upper plenum and the hot-leg piping system of JSFR was developed. Unsteady fluid flow analyses were then conducted by unsteady Reynolds averaged Navier-Stokes simulation (URANS) with Reynolds stress model. The appropriateness of the calculated results was discussed by comparing available scale model test results. Furthermore, a prototype model for vibration analysis of the hot-leg piping was developed. In the model, the transient pressure data predicted by the URANS were used as input data for the vibration analysis.
Tanaka, Masaaki; Takaya, Shigeru; Fujisaki, Tatsuya*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2014 Koen Rombunshu, p.203 - 204, 2014/09
As flow-induced vibration (FIV) in the hot-leg piping with the short elbow (curvature radius corresponds to diameter) is one of the targeted issues in this study, numerical estimation method for dynamic analysis of mechanical stress on piping has been developed. As the preliminary step in development of the fluid-structure mechanical interaction simulation method, time history response analysis of piping by using the time history data of fluid pressure obtained by the unsteady hydraulics simulation as its boundary condition was attempted. Through the numerical results, potential capability of the dynamic analysis of piping were confirmed and unsteady behavior of pipe due to the unsteady flow phenomena was analyzed.
short elbow piping systems of JSFRTanaka, Masaaki; Fujisaki, Tatsuya*; Oki, Hiroshi*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2013 Koen Rombunshu, p.141 - 142, 2013/09
Numerical simulation method to analyze unsteady flow phenomena in the upper plenum and the hot-leg piping system in Japan Sodium cooled Fast Reactor (JSFR) has been developed. A flow-induced vibration in the hot-leg piping is mainly targeted in this study. A numerical model of the hot-leg piping integrated into a full domain of the upper plenum was constructed. By comparing with the existing experimental results, applicability of numerical model was confirmed.
Tanaka, Masaaki; Ohshima, Hiroyuki; Yamano, Hidemasa; Aizawa, Kosuke; Fujisaki, Tatsuya*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1699 - 1708, 2010/06
Numerical simulations with the Unsteady Reynolds Averaged Navier Stokes equation (U-RANS) approach are examined at high Reynolds number conditions in the 1/3 scale water experiment which is a scale model of the hot-leg piping of JSFR. Applicability of the U-RANS approach to the unsteady flow through the short-elbow is confirmed and the pressure fluctuation generation mechanism in relation to the unsteady large scale eddy motion is clarified. Moreover, preliminary numerical simulation of a full-scale model of the hot-leg piping of JSFR is examined and applicability of the URANS approach to the hot-leg piping of JSFR is confirmed.
Tanaka, Masaaki; Ohshima, Hiroyuki; Yamano, Hidemasa; Aizawa, Kosuke; Fujisaki, Tatsuya*
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9285_1 - 9285_8, 2009/05
In this study, applicability of the Unsteady Reynolds Averaged Navier-Stokes equations (U-RANS) approach with the Reynolds Stress Model (RSM) to the flow through the short-elbow under high Reynolds number condition was examined through an analysis of 1/3 scaled water experiment (mean velocity is 9.2 m/s and Re number is 3.7E+6) simulates the hot-leg piping of the JSFR. Applicable numerical options in the U-RANS approach were identified through parametric study for time integration schemes and discretization schemes comparing with the experimental results. By the numerical simulation, a dominant frequency component of the pressure fluctuation was captured and mechanism of the pressure fluctuation generation was predicted in the relation of horseshoe eddy structure and secondary flow around the elbow outlet.
Saito, Kimiaki; Teshima, Naoya; Suzuki, Yoshio; Nakajima, Norihiro; Saito, Hidetoshi*; Kunieda, Etsuo*; Fujisaki, Tatsuya*
FUJITSU Famirikai Rombunshu (Internet), 20 Pages, 2009/02
We research and develop IMAGINE as the dose calculation system, which can calculate the dose distribution in the patient body very fast and with high precision by a super parallel calculation, for the contribution to QoS of the radiation therapy and the development of the eading-edge treatment technology. When IMAGINE system will be actually operated, it is anticipated that IMAGINE system accepts the large amount of requests of the dose calculation from several radiotherapy facilities at the same time. So, we entertained that the dose calculation was distributed by the gridization of IMAGINE system and the waiting time of its was minimized. In this article, we introduce the problem about the gridization of IMAGINE system, the solution to it and the method to easily enable the gridization of the application without the programming skill.
Saito, Kimiaki; Saito, Hidetoshi*; Kunieda, Etsuo*; Narita, Yuichiro*; Myojoyama, Atsushi*; Fujisaki, Tatsuya*; Kawase, Takatsugu*; Kaneko, Katsutaro*; Ozaki, Masahiro*; Deloar, H. M.*; et al.
Joho Shori, 48(10), p.1081 - 1088, 2007/10
no abstracts in English
Deloar, H. M.*; Kunieda, Etsuo*; Kawase, Takatsugu*; Tsunoo, Takanori*; Saito, Hidetoshi*; Ozaki, Masahiro*; Saito, Kimiaki; Takagi, Shunji*; Sato, Osamu*; Fujisaki, Tatsuya*; et al.
Medical Physics, 33(12), p.4635 - 4642, 2006/12
Times Cited Count：13 Percentile：39.28(Radiology, Nuclear Medicine & Medical Imaging)We are investigating three-dimensional converging stereotactic radiotherapy for small lung tumors with better dose homogeneity at the target. A computed tomography radiotherapy was simulated with BEAMnrc X-ray energy of 147.5, 200, 300, and 500 kilovoltage system was validated by comparing calculated and measured percentage of depth dose in a water phantom for the energy of 120 and 147.5 kVp. A thorax phantom and CT data from lung tumors compare dose homogeneities of kVp energies with MV energies of 4, 6, and 10 MV. Three non-coplanar arcs of the target were employed. The Monte Carlo dose data format was converted to the XiO RTP format to compare dose homogeneity, differential, and integral dose volume histograms of kVp and MV energies. In terms of dose homogeneity and DVHs, dose distributions at the target of all kVp energies with the thorax phantom were better than MV energies, with mean dose absorption at the ribs of 100%, 85%, 50%, 30% for 147.5, 200, 300, and 500 kVp, respectively. Considering dose distributions and reduction of the enhanced dose absorption at the ribs, a minimum of 500 kVp is suitable for the lung kVp 3DCSRT system.
