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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.
; ; ; Yamaguchi, Akira
JNC TN9400 2000-109, 96 Pages, 2000/11
JNC-TN9400-2000-109.pdf:9.56MB
Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiaied based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1)Ther is no remarkable difference betweeen liquid sodium and luquid Pb-Bi in characteristics of internal flows and free surface charatristics based on Fr number. (2)the AQUA-VOF code has a potentiall enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [thermal Stratification Phenomena] (1)On-set position of thermal entainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. 0n the other hand, the position in the case of C0
gas was shifted to upstream side with decreasing of Ri number. (2)Destruction speed of the thermal stratyification interface was dependent on thermal diffusivity as fluid properties. therefor it was concluded that an elimination method is necessary for the interface generated in C0 gas. [thermal Striping Phenomena] (1)Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2)To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it isnecessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phenomena] (1)Fundamental behavior of the natural convection in various coolant follows buoyant jet ....
; Yamaguchi, Akira
JNC TN9400 2000-056, 150 Pages, 2000/05
JNC-TN9400-2000-056.pdf:6.67MB
[Purpose] The work was performed to make clear thermal-hydraulic issues resulting in the use of various coolants for fast reactors. [Method] Plant design features due to a use of working fluid other than sodium and design concepts relating a simplification of safety related systems were investigated. And based on the results, quantitative evaluation was made on the topical themal-hydraulic issues. Then both thermal stratification and striping phenomena were evaluated by the used of thermo-hydraulics computer programs. [Results] (1)Thermal-hydraulic issues Topical thermal-hydraulic issues of gaseous and heavy metal cooled reactors were extracted. (a)Gas cooled reactors : natural circulation,flow-induced vibration, depressurization accident (b)Heavy metal cooled reactors : thermal stratification, flow-induced vibration, sloshing And also the thermal-hydraulic issues relating compact reactor assembly and RVACS were extracted resulting from a simplification of safety related systems. (2)Evaluation of thermal stratification and striping phenomena. The following order of affects for the phenomena was obtained: (a) Thermal stratification: CO 
Sodium Lead, (b) Thermal Striping: CO Lead Sodium
Yang Zumao*;
JNC TN9400 2000-009, 81 Pages, 2000/02
JNC-TN9400-2000-009.pdf:47.3MB
It is important to study thermal stratification and striping phenomena for they can induce thermal fatigue failure of structures. This presentation uses the AQUA code, which has been developed in Japan Nuclear Cycle Development Institute (JNC), to investigate the characteristics of these thermal phenomena in water, liquid sodium, liquid lead and carbon dioxide gas. There are altogether eight calculated cases with same Richardson number and initial inlet hot velocity in thermal stratification calculations, in which four cases have same velocity difference between inlet hot and cold fluid, the other four cases with same temperature difference. The calculated results show : (1) The fluid's properties and initial conditions have considerable effects on thermal stratification, which is decided by the combination of such as thermal conduction, viscous dissipation and buoyant force, etc., and (2) The gas has distinctive thermal stratification characteristics from those of liquid because for
Muramatsu, Toshiharu;
PNC TN9410 92-106, 354 Pages, 1992/04
A natural circulation analysis in the upper plenum of the MONJU reactor was conducted for transient simulating a pump coast down and reactor scram to a full-power operation condition using a multi-dimensional code AQUA. In the analysis, full options of the AQUA code (higher-order differencing schemes, an algebraic stress turbulence model, an adaptive Fuzzy control system, etc.) were used to obtain a refined numerical result. From the analysis, the following results have been obtained. (1)In a steady-state calculation simulating the full-power operation condition, maximum axial temperature gradient 154
C/m was calculated at the region between the upper and the lower flow holes. Therefore detailed measurements are necessary for thermal stress evaluation of internal components due to the axial temperature gradient at various power operation conditions. (2)In a transient caluculation simulating a natural circulation phenomenon, it was confirmed that a rising speed of the thermal stratification interface is delayed due to the decrease of a effective mixing volume in the upper plenum region. And the AQUA code calculated a discontinuity temperature transient (a hot shock continued from a cold shock) at the outlet nozzle of the reactor vessel due to the change of locally flow patterns in the upper plenum. Therefore it was concluded that detailed investigation is necessary using experimental data in various power operation conditions. (3)A gentle temperature transient was calculated with the AQUA code in comparison with a one-dimensional code. It is concluded that the one-dimensional code yields a conservative numerical result.
;
PNC TN9410 91-217, 65 Pages, 1991/07
For the purpose of the participation to a benchmark exercise of the 7th Meeting of the International Association for Hydraulic Research (IAHR) Working Group on Advanced Nuclear Reactors Thermal Hydraulics, which is to be held in Kernforschungszentrum Karlsruhe GmbH, FRG, August 27-29, 1991, two experimental cases of the benchmark problems were calculated using a combined method of higher-order accurate schemes and the Algebraic Stress turbulence Model (ASM) of a multi-dimensional themohydraulic analysis code AQUA developed at PNC. From the analyses, the following analytical results have been obtained: (1)Penetration phenomena at the inlet channel observed in experiments were predicted well with the combined method. (2)Axial distributions of temperature and horizontal velocity components in the test plenum agreed with experiments.
Muramatsu, Toshiharu
PNC TN9410 90-147, 115 Pages, 1990/10
In-vessel thermohydraulics analysis was conducted for transient simulating a pump coast down and reactor scram (manual reactor trip event) to evaluate effects of an inner barrel on a large fast breeder reactor. Then four thermohydraulics phenomena, a thermal stratification, a main loop temperature transient, a circumferential temperature distribution and a sodium surface velocity were discussed. Through the analysis using the multi-dimensional code AQUA and the discussion, the following have been effects of the inner barrel as obtained: [Thermal Stratification] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of an axial temperature distribution can be neglected from a structural design. [Main Loop Temperature Transient] An inner barrel is required. Because a cold shock with maximum temperature transient -2.0C/s occurred at a outlet nozzle when an inner barrel was not equipped. [Circumferential Temperature Distribution] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of the temperature distribution can be neglected from a structural design. But further investigation on a thermal stress at reactor vessel is necessary. [Sodium Surface velocity] An inner barrel is unnecessary. From the above results, it is concluded that an inner barrel is unnecessary if the cold shock is improved by a increase of effective mixing region on a design.
Muramatsu, Toshiharu; Ninokata, Hisashi
PNC TN9410 89-132, 59 Pages, 1989/09
For the purpose of the establishment of analytical model for turbulent flow behavior related to in-vessel thermal stratification phenomena, an algebraic stress turbulent model (ASM) has been implemented into AQUA in the place of the k-
turbulence model. The new turbulence model has provided high-accurate results in thermal stratification analysis due to the fact that empirical constants such as the turbulence prandtl number prt have been eliminated. From the analyses of water and sodium experiments using the new model, the following results have been obtained: [Water experiment] (1)Calculated speed of stratification interface rise agreed well with the experiment under the conditions that the internal sloshing behavior of the stratification interface was not observed. (2)While under the experimental conditions with the internal sloshing behavior being present, a calculated sloshing amplitude was slightly underestimated by the ASM. One of the reasons is considered to be that the model constants were derived for steady-state flows and not for transient turbulent flows. [Sodium experiment] In general, calculated speed of stratification interface rise has agreed well with the experiment. From the above results, it was confirmed that the algebraic stress turbulence model is superior to the conventional k- turbulence model on accuracy for an analysis of thermal stratification phenomenon.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
In the ULOHS tests performed in the experimental fast reactor U.S. EBR-II, the thermal stratification in the cold pool (CP) has influence on the whole plant behavior during the events because the secondary sodium pump tripped without scram nor tripping the primary pumps. In order to create the one-dimensional model for the CP of the plant dynamics analysis code, the multi-dimensional thermal hydraulics analyses using computational fluid dynamics (CFD) code were conducted to investigate the thermal hydraulics phenomena in the CP. It was found by comparison with the experimental data that the modeling of the detail sodium flow at the outlet of the intermediate heat exchanger, the leakage flow from the inner components to the cold pool, and the heat radiation from the CP to the atmosphere was important to the evaluation of the thermal stratification.