Refine your search:     
Report No.
Search Results: Records 1-14 displayed on this page of 14
  • 1

Presentation/Publication Type

Initialising ...


Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...


Initialising ...


Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Development of multi components and multi phase numerical method with chemical reaction; Examination of Multi phase numerical method

Takata, Takashi; Yamaguchi, Akira

JNC-TN9400 2001-018, 78 Pages, 2000/11


In the steam generator using liquid sodium, Water intensely reacts with sodium when it leaked out from a heat tube. It is important to evaluate an influence of the sodium-water reaction to, such as, heat tubes surrounding a leakage and the generator. In the past, evaluations of this phenomenon have been carried out by experiments. However it is difficult to extrapolate an effect by configuration of a heat tube or change of operating condition, etc. and experiments using sodium need incredible cost. Then quantification by a numerical method is desirable. To develop a multi component and multi phase numerical method with chemical reaction, fundamental models of a multi phase numerical method are selected with organizing previous works in this paper, as follows Fluid model : multi fluid model Pressure model : one pressure model Solving method : HSMAC (Highly Simplified Maker And Cell) method Two-dimensional two-phase flow analysis technique is developed to evaluate a validity of these models. And verification analyses are carried out shown in the following. (1)TWo-dimensional square cavity flow (2)Two-dimensional natural convection in a square cavity (3)Air blow down from a pressure vessel (4)Dam break-down problem (5)Edwards pipe blow down problem. In each verification analysis, good agreements are obtained and the validity of the models to a multi phase numerical method is confirmed.

JAEA Reports


; ; *; Yamaguchi, Akira

JNC-TN9400 2000-109, 96 Pages, 2000/11


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$$_{2}$$ 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$$_{2}$$ gas. [thermal Striping Phenomena] (1)Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO$$_{2}$$ 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 ....

JAEA Reports

Thermal-hydraulic of partially blocked fuel subassembly with porous media

; Ohshima, Hiroyuki; Yamaguchi, Akira

JNC-TN9400 2001-019, 97 Pages, 2000/10


The analysis code for investigations of local subassembly phenomena, which has been recognized as an issue of local subassembly accidents, has been required and developed at JNC. It is desirable for the analysis code to be applicable to various blockage conditions and random position of the blockage formation and to evaluate conservatively on the safety assessment with high accuracy, In this study, for the purpose of verifying the application and issues of the subchannel analysis code ASFRE-IV which evaluates thermal hydraulic phenomena in the porous blockage regions, the ASFRE-IV validation analysis was carried out on the basis of the data of an experiment investigation on a local porous blockage in a fuel subassembly performed by Reactor Engineering Groop, O-arai Engineering Center, JNC. Calculational results indicated that ASFRE-IV could reproduce the coolant temperature profile in a fuel subassembly and the peak temperature in the local subchannel conservatively.

JAEA Reports

Discretization of Convection-Diffusion Equations With Finite; Difference Scheme Derived From Simplified Analytical Solutions

Kriventsev, V.

JNC-TN9400 2000-094, 35 Pages, 2000/09


Most of thermal hydraulic processes in nuclear engineering can be described by general convection-diffusion equations that are often can be simulated numerically with finite-difference method(FDM). An effective scheme for finite-difference discretization of such equations is presented in this report. The derivation of this scheme is based on analytical solutions of a simplified one-dimensional equation written for every control volume of the finite-difference mesh. These analytical solutions are constructed using linearized representations of both diffusion coefficient and source ter. As a result, the Efficient Finite-Differencing (EFD) scheme makes it possible to significantly improve the accuracy of numerical method even using mesh systems with fewer grid nodes that, in turn, allows to speed-up numerical simulation. EFD has been carefully verified on the series of sample problems for which either analytical or very precise numerical solutions can be found. EFD has been compared with other popular FDM schemes including novel, accurate (as well as sophisticated) methods. Among the methods compared were well-known central difference scheme, upwind scheme, exponential differencing and hybrid schemes of Spalding. Also, newly developed finite-difference schemes, such as the quadratic upstream (QUICK) scheme of Leonard, the locally analytic differencing(LOAD) scheme of Wong and Raithby, the flux-spline scheme proposed by Varejago and Patankar as well as the latest LENS discretization of Sakai have been compared. Detailed results of this comparison are given in this report. These tests have shown a high efficiency of the EFD scheme. For most of sample problems considered EFD has demonstrated the numerical error that appeared to be in orders of magnitude lower than that of other discretization methods. 0r, in other words, EFD has predicted numerical solution with the same given numerical error but using much fewer grid nodes. In this report, the detailed ....

JAEA Reports

Survey of thermal-hydraulic correlations for gas, lead and lead-bismuth coolants

; Ohshima, Hiroyuki

JNC-TN9400 2000-078, 130 Pages, 2000/06


Feasibility study is being carried out at JNC to generate new concepts of practical fast breeder reactors. ln this study, appropriate thermal-hydraulic correlations for several kinds of coolants are required to assess thermal-hydraulics of proposed core/fuel-assembly designs, which have different characteristics from traditional liquid-sodium cooled fast reactors, e.g., ribbed fuel pins and fuel pin square arrangement with spacer. ln the present report thermal-hydraulic correlations for carbon di-oxide, helium, lead, and lead-bismuth cooled reactors were surveyed. Several candidates for pressure drop coefficient and heat transfer coefficient of each coolant were picked from available papers and literatures and were examined by using the design specifications of ETGBR (carbon di-oxide cooled reactor), GBR4(helium cooled reactor) and BREST300 (lead, lead-bismuth cooled reactor) as well as existing experimental data. Finally thermal-hydraulic correlations of each coolant, which are applicable to the regions from laminar to turbulent flow, were proposed.

JAEA Reports

Development and validation of Multi-DimensionaI sodium combustion analysis code AQUA-SF

Takata, Takashi; Yamaguchi, Akira

JNC-TN9400 2000-065, 152 Pages, 2000/06


ln the liquid metal fast reactor (LMFR) using liquid sodium as a coolant, it is important to evaluate the effect of the sodium combustion on the structure, etc. Most of the previous analytical works are based on a zone model, in which the principal variables are treated as volume-average quantities. Therefore spatial distribution of gas and structure temperatures, chemical species concentration are neglected. Therefore, a multi-dimensional sodium combustion analysis code AQUA-SF (Advanced simulation using Quadratic Upstream differencing Algorithm - Sodium Fire version) has been developed for the purpose of analyzing the sodium combustion phenomenon considering the multi-dimensional effect. This code is based on a multi-dimensional thermal hydraulics code AQUA that employs SIMPLEST-ANL method. Sodium combustion models are coupled with AQUA; one is a liquid droplet model for spray combustion, and the other is a flame sheet model for pool combustion. A gas radiation model is added for radiation heat transfer. Some other models necessary for the sodium combustion analysis, such as a chemical species transfer, a compressibility, are also added. ln AQUA-SF code, bounded QUICK method in space scheme and bounded three-point implicit method in time scheme are implemented. Verification analyses of sodium combustion tests shown in the following have been carried out. (1)pool combustion test (RUN-D1) (2)spray combustion test (RUN-E1) (3)sodium leakage combustion test (Sodium Fire Test-II) (4)smaII-scale leakage combustion test (RUN,F7-1) ln each verification analysis, good agreements are obtained and the validity of AQUA-SF code is confirmed.

JAEA Reports

lnvestigation of thermal-hydraulic issues resulting in the use of various coolants

; Yamaguchi, Akira

JNC-TN9400 2000-056, 150 Pages, 2000/05


[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$$_{2}$$ $$<$$ Sodium $$<$$ Lead, (b) Thermal Striping: CO$$_{2}$$ $$<$$ Lead $$<$$ Sodium

JAEA Reports

Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor; Comparisons of the decay heat removal characteristics on Lead, Lead-Bismuth and Sodium cooled reactors

; *; Ohshima, Hiroyuki; Yamaguchi, Akira

JNC-TN9400 2000-033, 94 Pages, 2000/04


The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.

JAEA Reports

Experimental study on the avoidance and suppression criteria for the vortex-induced vibration of a circular cylinder

; ; ;

JNC-TN9400 2000-012, 43 Pages, 2000/03


Experimental validation of the design method to prevent the failure of a thermometer well by the vortex-induced vibration has been performed for the effect of structure damping of a cylinder. The available experimental data in piping were very limited for the high damping region in water flow, because of the difficulty to increase the structure dumping for the one-side supported cylinder in experiments. ln this experiment, high viscosity fluid was charged into the tested cylinders to control the cylinder's damping. Resulting values of the reduced damping are 0.49, 0.96, 1.23, 1.98, 2.22 in the experiments. Reduced velocity(Vr) was increased gradually in the range of 0.7 $$leq$$ Vr $$leq$$ 5(Reynolds number at Vr=1 is 8$$times$$10$$^{4}$$). The displacements of the cylinder by the vortex-induced vibration were measured. As the results, Tested cylinders of reduced damping 0.49 and 0.96 showed vortex-induced vibration in flow direction at Vr $$>$$ 1 region. However, in case of reduced damping of 1.23, 1.98 and 2.22, the vortex-induced vibrations in flow direction were suppressed lower than the 1% displacement of the cylinder diameter. ln conclusion, it is confirmed that the suppression criteria of the "Vr $$<$$ 3.3 and Cn $$>$$ 1.2" for vortex-induced vibration in flow direction, which is used in ASME code; "Boiler and Pressure VesseI Code Sec.III Appendix N-1300" and the "FIV design guide in JNC", is reasonably applicable to the one-side supported cylinder in water flow piping.

JAEA Reports

Investigation of in-vessel thermal stratification characteristics for large scale liquid sodium cooled fast breeder reactors with general purpose multi-dimensional code AQUA

Hendro Tjahijono*;

JNC-TN9400 2001-023, 90 Pages, 2000/02


Thermal stratification phenomena is an important problem in liquid metal fast breeder reactor (LMFBR) vessel where it can induce thermal fatigue failure of structures. The phenomena is observed in the upper plenum zone under reactor scram conditions. Investigation on characteristics of this phenomena is performed using a general purpose multi-dimensional code AQUA. There are 3 sizes of FBR vessel having been investigated such as: large size with 9.6 m of diameter, medium size with 8.16 m of diameter and small size with 6.72 m of diameter. Two types of calculation have been applied in this investigation such as: steady state calculation with QUICK / FRAM and transient calculation with QUICK / FRAM and first order upwind. In this calculation, fully implicit method has been used for the beginning followed by semi-implicit method until steady state condition and then transient run during 1000 seconds simulating a scram condition from full power operation condion. For large size of FBR, three dimensional (3D) model and two dimensional (2D) model have been performed. Due to the very long CPU time in 3D model, for medium and small size, only 2D model has been applied. Thermal stratification is investigated after reactor scram in the both model. The average rising speed of the thermal stratification calculated using least square method in the period between 100s to 1000s after scram are 3.96 m/hr in small vessel, 8.28 m/hr in medium vessel, 6.84 m/hr in large vessel 2D model and 7.56 m/hr in large vessel 3D model calculated after 200s. The maximum of the maximum axial temperature gradients in all cases during transient are around : 800 $$^{circ}$$C/m in small vesse1, 700$$^{circ}$$C/m in medium vessel, 1000 $$^{circ}$$C/m in large vessel 2D model and 400$$^{circ}$$C/m in large vessel 3D model.

JAEA Reports

Current status and future plan for thermaI striping investigations at JNC

; kasahara, Naoto; ; ; Kamide, Hideki

JNC-TN9400 2000-010, 168 Pages, 2000/02


Thermal striping is significant issue of the structural integrity, where the hot and cold fluids give high cycle fatigue to the structure through the thermal stress resulted from the time change of temperatur distibution in the structure. In the sodium cooled fast reactor, temperature change in fluid easily transfers to the structure because of the high thermal conductivity of the sodium. It means that we have to take care of thermal striping, The thermal striping is complex phenomena between the fluid and structure engineering fields. The investigations of thermal striping are not enough to evaluate the integrity directly. That is the fluctuation intensity at the structure surface is assumed to be temperature difference between source fluids (upstream to the mixing region) as the maximum value in the design. 0therwise, the design conditions are defined by using a mockup experiment and large margin of temperature fluctuation intensity. Furthermore, such evaluation manners have not yet been considered as a design rule. Transfer mechanism of temperature fluctuation from fluid to structure has been investigated by the authors on the view points of the fluid and structure. Attenuation of temperature fluctuation was recognized as a dominant factor of thermal fatigue. We have devdoped a numerical analysis system which can evaluate thermal fatigue and crack growth with consideration of the attenuation of temperature fluctuation in fluid, heat transfer, and structure. This system was applied to a real reactor and the applicability was confirmed. Further verification is planned to generalize the system. For the higher cost performance of the fast reactor, an evaluation rule is needed, which can estimate thermal loading with attenuation and can be applied to the design. An idea of the rule is proposed here. Two methods should be prepared; one is a precise evaluation method where mechanism of attenuation is modeled, and the other is simple evaluation method where ...

JAEA Reports

Numerical Investigation on Thermal Stratification and Striping Phenomena in Various Coolants

Yang Zumao*;

JNC-TN9400 2000-009, 81 Pages, 2000/02


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

JAEA Reports

Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes (IV); Investigation on second-order moments in coolant mixing region

JNC-TN9400 2000-008, 323 Pages, 2000/02


This rport presents numerical results on theemal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm$$^{I.D.}$$) with a 90$$^{circ}$$ elbow and a branch pipe, and four parameters, j.e., (1)diameter ratio $$alpha$$ between both the pipes, (2)flow velocity ratio $$beta$$ between both the pipes, (3)angle $$gamma$$ between both the pipes, and (4)Reynolds number Re. From the numerical investigations, the following characteristics were obtained: (1)According to the decreasing of the diameter ratio, significant area of second-order moments was expanded in the fixed condition of $$beta$$=1.0. (2)Significant second-order moments area was expanded for the increasing of the flow velocity ratio $$beta$$ specified by varying of the main pipe velocity in the case of a $$alpha$$ = 1.0 constant condition. 0n the other hand, the area was expanded for the decreasing of the velocity ratio $$beta$$ defined by varying of the branch pipe velocity in the case of a $$alpha$$ = 3.0 constant condition. (3)Maximum second-order moments values were generated in the case of $$gamma$$ = 180$$^{circ}$$ due to the influence of interactions between main pipe flows and jet flows from the branch pipe. (4)According to the increase of Reynolds number, significant area of second-order moments was expanded due to the activation of turbulence mixing in the main pipe.

JAEA Reports

Thermal-Hydraulic investigation on severaI fast reactor design concepts

Ohshima, Hiroyuki; ; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *

JNC-TN9400 2000-077, 223 Pages, 1999/05


The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.5$$sim$$0.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...

14 (Records 1-14 displayed on this page)
  • 1