Development of computer program for detailed thermal-hydraulic analysis in a fast reactor fuel subassembly (2); Incorporation of turbulence models

Ohshima, Hiroyuki; Ohshima, Hiroyuki; Imai, Yasutomo*

JNC TN9400 2003-045, 74 Pages, 2003/06

JNC-TN9400-2003-045.pdf:3.05MB

As a thermal-hydraulic evaluation tool for fast reactor fuel subassemblies with high burn-up ratio, a numerical analysis system in which a subchannel analysis program and a detailed thermal hydraulic analysis program are utilized interractively is under development. This system enables us not only to clarify thermal hydraulic characteristics that cannot be revealed by experiments due to measurement difficulty but also to contribute to the rational safety design and assessment. This report describes the incorporation of turbulence models to the detailed thermal hydraulic analysis program SPIRAL-II and its verification study. In addition to the standard k- two-equation model, Renormalization Group(RNG) k- model and Algebraic Stress Model(ASM) were incorporated to SPIRAL-II as turbulence models. The former utilizes fewer empirical constants than the standard k- model and is believed to be more accurate especially in low Reynolds number regions. The latter treats Reynolds stresses directly and therefore has applicability to anisotropic turbulent flow. It is common for three models that similar transport equations for turbulent kinetic energy k and dissipation rate are discretized and solved by galerkin method. With respect to ASM, Daly-Harlow model is applied to the diffusion terms of the transport equations and algebraic expressions related to Reynolds stresses are solved by Newton-Raphson method using calculated k and . A wall function (Reichardt function) is applied to each model as the boundary condition treatment. A verification study of improved SPIRAL-II was carried out using three kinds of problems: turbulent flow between parallel walls, backstep facing turbulent and turbulent flow in a square duct. From these calculation results, the validity of the improved program was confirmed and prediction characteristics of each turbulent model were clarified.

Modification of multi-dimensional sodium-water reaction analysis code: SERAPHIM and sensitivity analyses on the early stage of leakage

Takata, Takashi; Yamaguchi, Akira; Watanabe, Akira*

JNC TN9400 2003-024, 129 Pages, 2003/04

JNC-TN9400-2003-024.pdf:9.72MB

Modification of the SERAPHIM code, which was developed to deal with the sodium-water reaction phenomena in a steam generator of the LMFR, has been carried out With regard to the reaction mechanism, the surface reaction model is improved. The hydrated effect on the reaction rate in the gas-phase reaction is estimated and the reaction product in the sodium-water reaction is investigated. The SERAPHIM code is also parallelized using the MPI (Message Passing Interface) method to apply to a large-scale simulation such as over one million nodes with a limited PC resource. The analysis region and the memory needed in the analysis are discrete in the SERAPHIM code. The execution time becomes approximately l7 times faster in case of 16 CPUs. The calculation of one million nodes can be available even in a PC cluster system with 1 GB memory. Sensitivity analyses of the surface reaction model in the beginning of the water vapor leakage are carried out and the following conclusion are obtained. 1)Gas region develops quickly in case of no chemical reaction. The pressure peak is obtained at an early stage of the leakage regardless of the chemical reaction. 2)The difference of the initial pressure strongly affects the development of the gas region and the flow-field. However,It is less influential on the maximum temperature. The maximum temperature depends not only on the chemical reaction characteristics such as the gas region and heat generation but also on the thermal-hydraulics characteristics in the multi-phase flow field. 3)Considering the evaporation of sodium hydroxide (NaOH), the maximum gas temperature of approximately 1200deg is observed in the analysis. In the standard state, the chemical reaction that produces a gas state sodium oxide and hydrogen gas from liquid sodium and water vapor is endothermic. In is concluded that the evaporation of sodium hydroxide is very important phenomena in the sodium-water reaction.

Analytical study on thermal-hydraulics of gas cooled fast reactors

Ohshima, Hiroyuki; Nishimura, Motohiko*

JNC TN9400 2002-072, 97 Pages, 2002/08

JNC-TN9400-2002-072.pdf:3.85MB

A feasibHity study has been carried out at JNC to construct new design concepts of commercialized fast reactors. This report describes two kinds of numerical investigations related to thermal-hydraulics of gas-cooled fast reactors of which design studies are being performed as part of the feasibility study. A series of thermal hydraulic analyses was carried out using multi-dimensional analysis program AQUA in order to confirm the heat removal capability of the helium-gas-cooled fast reactor with a coated-particle-type fuel assembly design under a rated power operation, a low power/low flow and an accident conditions. The calculation results indicates that the lateral gas flow which is indispensable for normal heat removal in the fuel particle region is kept under each calculation condition and the maximum temperature does not exceed the tentative design limitation as far as the inlet surface permeability is uniform. Only in the case of high pressure and very low flow rate, a possibility that local high temperature region exceeding the design limitation appears may not be denied due to upward flow driven by buoyancy force. Improvement on knowledge of gas property functions in high temperature and pressure drop correlations are required for more accurate analysis. Natural circulation decay heat removal characteristics of the CO gas cooled fast reactor are examined using a one-dimensional nuclear-thermal-hydraulics network analysis code, MR-X equipped with correlations for the core thermal-hydraulics of gas cooled fast reactors. Simulation parameters are the shutdown time of steam generators (SGs), restart time of gravitational water feed to the SGs, and flow rate of the feed water. It was predicted that the reactor satisfied limitation of the maximum cladding temperature under the realistic operation condition of SGs with the shutdown time of 30 s and the restart time of 20 min. Moreover, the cladding temperature still satisfied the limitation, even ...

None

Muramatsu, Toshiharu; *; Yamaguchi, Akira

JNC TN9400 2002-022, 96 Pages, 2002/06

None

Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes (VI); Numerical evaluations of arched-vortex characteristics in non-isothermal fields

*;

JNC TN9400 2002-011, 37 Pages, 2002/05

JNC-TN9400-2002-011.pdf:5.36MB

Numerical analyses for turbulence thermal mixing, the aim of which is to evaluate relationship between hydrodynamics and temperature distribution of an archedvortex, were carried out using the direct numerical simulation code DINUS-3. From the analyses, the following results have been obtained: (1)Transportation period of the arched-vortex and distance between the arched--vortices were kept constant in isothermal and non-isothermal conditions. (2)The transportation period of afched-vortex was decreased with increasing Reynolds number under the condition of the constant flow velocity ratio between both coolant pipes. (3)One of the main reasons for this behavior was considered that the motion of the cold fluid flowing out of the branch pipe was restricted by the difference of fluid density between the branch and the main pipes. The amplitudes of the cross flow velocity fluctuation in the leg region of the arched-vortex were larger than those under isothermal condition. (4)It was confirmed that the arched-vortex consists of two kinds of vortexes, i.e., a longitudinal vortex generated by a shear motion at the top of the arched-vortex, and a horizontal vortex by shedding motion at both sides of the branch jet flow.

Study on in-vessel Thermohydralics phenomena of sodium-cooled fast reactors (I); Numerical investigation for the Rationalization of Hydrodynamics in the upper plenuln

; *; Yamaguchi, Akira

JNC TN9400 2001-117, 60 Pages, 2002/02

JNC-TN9400-2001-117.pdf:7.06MB

A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possiblity. From the analysis, the following results were obtained. (1)It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2)Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3)Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4)Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5)Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1)Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2)Alleviation measures of vortex concentration at free surface. (3)Separation measures of 3-dimensional vortex distributi

Thermal-hydraulic analysis in local fuel region of helium gas-cooled fast breeder reactor with coated-particle-type fuel

; Ohshima, Hiroyuki

JNC TN9400 2001-101, 85 Pages, 2001/06

JNC-TN9400-2001-101.pdf:2.53MB

Feasibility Study is being carried out at JNC to generate new concepts for commercialized fast breeder reactors. In this study, a helium gas-cooled reactor with coated-particle-type fuel is proposed, as one of the candidates for fast breeder reactors. Each fuel assembly has a compartment of an annular duct shape and the annular space of the compartment is filled with coated-particle-type fuels. The assessment of heat removal capability of coolant flowing in the coated-particle-fuel region and the endurance of fuel particle is one of the important issues in the reactor safety. In the present study, a thermal-hydraulic analysis was carried out in order to clarify flow and temperature fields in a local coated-particle-fuel region as well as in-particle temperature distributions. The FLUENT code was applied to this numerical analysis and the simulations were performed using five face-center cubic unit cells, which were combined with one another in the flow direction. Through the analysis, it was confirmed that the extreme temperature peak in coolant did not appear in the local Coated-particle-fuel region and the temperature in a coated fuel particle rises along the flow direction almost linearly except fuel core region. With respect to the surface temperature of a coated fuel particle, the maximum and the minimum temperatures appear at the downstream and the upstream contact points with neighboring particles, respectively. Further, the calculation results by FLUENT were compared with Ergun's correlation in order to verify the applicability of it to the pressure drop estimation in the coated-particle-fuel region. The friction coefficient estimated by FLUENT agreed with that by Ergun's correlation with errors from -11% to 20% for 2 Re 154.

MSV:Multi-Scaler Viscosity Model of Turbulence

Kriventsev, V.

JNC TN9400 2001-053, 38 Pages, 2001/04

JNC-TN9400-2001-053.pdf:1.31MB

Multi-Scale Viscosity (MSV) model is proposed for estimation of the Reynolds stresses in turbulent fully-developed flow in a wall-bounded straight channel of an arbitrary shape. We assume that flow in an "ideal" channel is always stable, i.e.laminar, but turbulence is developing process of external perturbations cased by wall roughness and other factors. We also assume that real flows are always affected by perturbations of any scale lower than the size of the channel. The turbulence can be modeled in form of internal or "turbulent" viscosity increase. The main idea of MSV can be expressed in the following phenomenological rule: A local deformation of axial velocity can generate the turbulence with the intensity that keeps the value of local turbulent Reynolds number below some critical value. Here, local turbulent Reynolds number can be defined in two different ways: (1)as a product of value of axial velocity deformation for a given scale and generic length of this scale divided by accumulated value of laminar and turbulent viscosity of lower scales (2)as a ratio of the difference between total kinetic energy and "flat-profile" kinetic energy to the work of friction forces In MSV, the only empirical parameter is the critical Reynolds number that is estimated to be around 100 in the former case and about 8.33 in the later. MSV model has been applied to the fully-developed turbulent flows in straight channels such as a circular tube and annular channel. Friction factor and velocity profiles predicted with MSV are in a good agreement with numerous experimental data. The MSV model can be classified as "zero-order" integral model of turbulence. Because of simplicity, MSV can be easily implemented for calculation of fuIly-developed turbulent flows in straight channels of arbitrary shapes including fuel assemblies of nuclear reactors. The intent of this report is to summarize the progress made in the development of the model of turbulence. Since the final ...