An application of a higher order finite difference method to a natural convection experiment in the hot plenum of an LMFBR

A two dimensional thermal-hydraulic analysis of a natural circulation experiment has been performed to evaluate the effectiveness of a higher order finite difference method for solving the Navier-Stokes and the energy equations. In the method, the convection terms appearing in each equation are solved by the Method of Characteristics using the third order Lagrange type polynomial as the interpolation function, and an iterative procedure is applied to solve the time derivative terms of each equation stably with second order accuracy. The analytical results have been compared with an experiment in which the temperature and the velocity distributions in the plenum region were measured with their fluctuations, and it was shown that the higher order finite difference method could simulate natural convection phenomena involving fluctuations well.     

Onset of water accumulation in the upper plenum with a perforated plate

This paper describes an experimental study on the onset of water accumulation above a perforated plate in a vertical air-water countercurrent flow. We experimentally investigate the effects of holes with the relatively large diameter (5 cm), the number of holes (4 holes and 12 holes), the thickness of plate (1 cm and 4 cm), asymmetric water injection, and the location of the air vent line on the onset of water accumulation. The present results indicate that the onset of water accumulation is promoted by a smaller diameter of hole, a smaller number of holes, and symmetric water injection whereas it has negligible dependence on the location of air vent line in the upper plenum. It turns out that the nondimensional superficial velocity, , fits the data better than either the Kutateladze number, K_k, or the interpolation parameter, . We develop a Wallis-type correlation for the onset of water accumulation when the diameter of the holes ranges between 0.7 and 5 cm, and the number of holes is greater than six.     

中国实验快堆热钠池热分层现象研究

中国实验快堆(CEFR)在紧急停堆工况下,会在热钠池上部空间形成热分层现象。热分层出现后,由于上腔室底部存在大量的冷钠(相对而言),这将延缓一回路自然循环的建立。同时,冷钠的存在还会降低自然循环的流量,并对事故停堆后堆芯的冷却产生不利影响。因此,热分层现象应当引起广泛注意。从设备结构的完整性分析上看,快堆热分层现象的出现对堆容器和部分堆内构件是不利的,会使这些部件在结构内部形成明显的热应力,对堆的安全运行构成隐患。本文调研了国内外在该领域的研究状况,分析国外已有的实验研究和理论计算进展,并结合快堆现有的计算分析程序,对CEFR的热分层现象进行深入和较为全面的计算分析。通过计算分析可以看到,在全厂断电工况下,在热钠池的上部会初步形成稳定的热分层,分层界面位于中间热交换器入口的下方,但是热分层现象不会对堆的自然循环构成影响。     

Measurement of turbulent flow phenomena for the lower plenum of a prismatic gas-cooled reactor

Mean velocity field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics design (Gas-Turbine-Modular Helium Reactor). The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered as a benchmark for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The primary objective of this paper is to document the experiment and present a sample of the data set that has been established for this standard problem.Present results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined crossflow—with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive index of the mineral oil working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3D) particle image velocimetry (PIV) system was used to collect the data. Inlet-jet Reynolds numbers (based on the hydraulic diameter of the jet and the time-mean average flow rate) are approximately 4300 and 12,400. Uncertainty analysis and a discussion of the standard problem are included.The measurements reveal complicated flow patterns that include several large recirculation zones, reverse flow near the simulated reflector wall, recirculation zones in the upper portion of the plenum and complex flow patterns around the support posts. Data include three-dimensional PIV images of flow planes, data displays along the coordinate planes (slices) and presentations that describe the component flows at specific regions in the model.     

Modeling strategies for unsteady turbulent flows in the lower plenum of the VHTR

Validation simulations are presented for turbulent flow in a staggered tube bank, geometry similar to the lower plenum of a gas-cooled high temperature reactor. Steady 2D RANS results are compared to unsteady 2D RANS results and experiment. The unsteady calculations account for the fact that nonturbulent fluctuations (due to vortex-shedding) are present in the flow. The unsteady computations are shown to predict the mean variables and the total shear stress quite well. Previous workers’ results indicate that 3D simulations are needed to obtain reasonable agreement; present results indicate 2D is sufficient. Best practices are based on requirements for the ASME Journal of Fluids Engineering.     

Optimization of the upper plenum of a PBMR to enhance thermal performance in the reactor core

An upper plenum of a PBMR type gas cooled nuclear reactor has been optimized using three-dimensional Reynolds-averaged Navier–Stokes (RANS) analysis and surrogate modeling technique. Shear stress transport turbulence model is used as a turbulence closure. Two geometric design variables viz., ratio of height of upper plenum to diameter of rising channels, and ratio of height of the slot at inlet to that at outlet, are used as design variables for the optimization. Design points are selected by Latin-hypercube sampling. The objective function is defined as a linear combination of uniformity of temperature distribution in the core and pressure drop through the upper plenum. The optimal point is determined through surrogate-based optimization method which uses RANS derived calculations at design points. The results show that the optimization improves considerably both the temperature uniformity and the friction performance.     

Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR

The pressure drop and heat transfer characteristics of wire-wrapped 19-pin rod bundles in a nuclear reactor subassembly of liquid metal cooled fast breeder reactor (LMFBR) have been investigated through three-dimensional turbulent flow simulations. The predicted results of eddy viscosity based turbulence models (k-?, k-ω) and the Reynolds stress model are compared with those of experimental correlations for friction factor and Nusselt number. The Re is varied between 50,000 and 150,000 and the ratio of helical pitch of wire wrap to the rod diameter is varied from 15 to 45. All the three turbulence models considered yield similar results. The friction factor increases with reduction in the wire-wrap pitch while the heat transfer coefficient remains almost unaltered. However, reduction in the wire-wrap pitch also enhances the transverse flow velocity in the cross-sectional plane as well as the local turbulence intensity, thereby improving the thermal mixing of coolant. Consequently, the presence of wire wrap reduces temperature variation within each section of the subassembly. The associated reduction in differential thermal expansion of rods is expected to improve the structural integrity of the fuel subassembly.     

PIV measurements of turbulent jet and pool mixing produced by a steam jet discharge in a subcooled water pool

This experimental research is on the fluid-dynamic features produced by a steam injection into a subcooled water pool. The relevant phenomena could often be encountered in water cooled nuclear power plants. Two major topics, a turbulent jet and the internal circulation produced by a steam injection, were investigated separately using a particle image velocimetry (PIV) as a non-intrusive optical measurement technique. Physical domains of both experiments have a two-dimensional axi-symmetric geometry of which the boundary and initial conditions can be readily and well defined. The turbulent jet experiments with the upward discharging configuration provide the parametric values for quantitatively describing a turbulent jet such as the self-similar velocity profile, central velocity decay, spreading rate, etc. And in the internal circulation experiments with the downward discharging configuration, typical flow patterns in a whole pool region are measured in detail, which reveals both the local and macroscopic characteristics of the mixing behavior in a pool. This quantitative data on the condensing jet-induced mixing behavior in a pool could be utilized as benchmarking for a CFD simulation of relevant phenomena.     

CFD analysis of a turbulent jet behavior induced by a steam jet discharge through a single hole in a subcooled water tank

A computational fluid dynamics (CFD) analysis for a turbulent jet flow induced by a steam jet discharged into a subcooled water pool was performed for 10 s of transients to investigate whether the currently available CFD codes can be suitably used as a tool to investigate the applicability of the existing semi-analytical correlations to a condensing jet-induced turbulent jet and to analyze the thermal-hydraulic behavior, such as global circulation and local hot spot, in a condensation pool for advanced light water reactors. As for the numerical experiment, a series of sensitivity calculations was conducted systematically to elucidate the major factors which can cause different analysis results by varying the mesh distributions, numerical models for a convection term and an eddy viscosity term. The effect of a difference in the velocity and the temperature distribution in a region between the sparger and the pool wall has not been observed in the afore-mentioned sensitivity calculations. The comparison of the CFD results with the test data shows that the CFD analysis does not accurately simulate the local phenomenon of a turbulent jet existing downstream of a steam jet. It was found that the value of the turbulent intensity at the inlet of the turbulent jet region is the most important factor because it can determine the boundary of a turbulent jet through a momentum diffusion process in a radial direction. The comparison of the CFD results with the test data shows that the CFD analysis can accurately simulate the local phenomenon of a turbulent jet existing downstream of a steam jet only when the CFD analysis reflects the physics of a turbulent jet.     

Establishment of the design requirements for a flow blockage detection system through a LES analysis of the temperature fluctuation in the upper plenum

When an assembly in a core is partially blocked, the temperature in the upper plenum fluctuates at an early stage. Therefore, the temperature fluctuation in the upper plenum can provide the information about a local blockage of an assembly. For developing the detection algorithm for the partial blockage, we analyzed the temperature fluctuation in the upper plenum due to the partial blockage in an assembly. The LES turbulence model in the CFX code was used for analyzing the temperature fluctuation in the upper plenum because the LES is suitable for analyzing the time dependent turbulence variables. After analyzing the temperature fluctuations in the upper plenum, we established basic design requirements for the flow blockage detection system through a FFT analysis and some statistical analysis. We concluded that response time of a measuring device was less than 13 m s and that it should cover a high temperature range of 1000 K. In addition, the resolution of the thermocouple was less than 2 K and its location should be within 25 cm from the exit of each assembly.     

Application of a coupled euler-lagrange computer program to the structural response of an LMFBR

This paper discusses the application of PISCES 2DELK, a coupled Euler-Lagrange computer code, to the response of an LMFBR to a hypothetical core-disruptive accident. The coupled Euler-Lagrange method allows one to use an Euler finite difference mesh to compute fluid and a Lagrange mesh to compute structural response. The finite difference equations are discussed, and a sample calculation is presented in which the sodium gas bubble expands around a movable internal structure.     

On the potential for plugged core pressurization following a hypothetical accident in an LMFBR

Steady state and transient conditions for a two-phase liquid-vapor cylindrical UO₂ pool are analytically investigated. The analysis relies on extrapolation of experimental results from internally heated boiling simulant fluid pools. The internal heat generation rates that can sustain a vapor-liquid pool under steady state consitions are calculated. The pressure transients that can be expected in the event of an imbalance between heat generation rate and heat removal rate are also calculated. The analysis presented demonstrates the significance of accounting for heat removal from the pool for proper assessment of its pressurization potential.     

Thermal fluctuations in the lower plenum of an high temperature reactor

This paper describes a typical study on thermal hydraulic problems on high temperature reactors. It deals with thermal stresses on the core outlet region of a new concept of high temperature reactor. The simulations point out the thermal fluctuations in the nominal state in the fluid and in the solid. First results are presented. They illustrate the complexity of the calculation due to particular geometry and boundary conditions. Qualitative analyses of the simulations reinforce the former evaluations on the oscillating character of the flow, the effects of mixing of different flows, and the consequences on the thermal load on the solid structures. In the future quantitative results can be used as source term for studies of solid mechanics. These calculations need also the computation of the global behaviour of the circuit. Simulations are performed with the TRIO_U/PRICELES code for the 3D-analysis and the CATHARE code for the system modelling. Both are developed by the CEA.     

Preliminary tests of particle image velocimetry for the upper plenum of a scaled model of a very high temperature gas cooled reactor

The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. During the design process multiple studies have been performed to develop safety codes for the reactor. Two major accidents of interest are the Pressurized Conduction Cooldown (PCC), and the Depressurized Conduction Cooldown (DCC) scenario. Both involve loss of forced coolant to the core, except the latter involves a pressure loss in the main coolant loop. During normal operation a circulator pumps the coolant into the upper plenum and down through the core, but following a loss of forced coolant the natural convection causes the flow to reverse to go through the core into the upper plenum. Computer codes may be developed to simulate the phenomenon that occurs in a PCC or DCC scenario, but benchmark data is needed to validate the simulations; previously there were no experimental test facilities to provide this. This study will cover the design, construction, and preliminary testing of a 1/16th scaled model of a VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets; the turbulent mixing of the jets was observed. After performing a sensitivity analysis the flow rate of a single pipe was extracted from the PIV flow field, and compared with an ultrasonic flowmeter and analytic flow rate. All the values lied within the uncertainty ranges, validating the test results.     

Evaluation of plutonium oxide aerosol release from an LMFBR in a hypothetical accident

This work was undertaken to prepare a computer code for the hazard evaluation of plutonium oxide aerosol released to the atmosphere in the event of a hypothetical accident in a 50 MW(th) scale LMFBR. The reactor building structure consists of semi-double containments as follows: the primary containment has a large volume in comparison with the secondary annular containment in which a part is connected to the atmosphere through an emergency filter system. Sodium oxide aerosol containing PuO₂---UO₂ fuel, fission products and structural steel agglomerates quickly by coagulation due to its high concentration. Simultaneously, the aerosol concentration decreases due to settling, plating and thermophoresis. Using the present code, the amount of PuO₂ aerosol leakage to the atmosphere was evaluated.     

Fragmentation of a molten corium jet falling into water

During a severe accident of a pressurized water nuclear reactor, a large mass of corium could pour into the vessel bottom as a compact jet. When the corium mass reaches the water at the bottom of the vessel, an intense fragmentation may occur. This could lead to a significant mixing of corium and water, likely to cause a steam explosion which could damage the structures. An analytical study has been established in order to quantify the corium jet fragmentation. This study consists mainly in modeling the vapor flow surrounding the jet as well as the instability which occurs at its interface. In comparison with previous studies, this model pays particular attention to the jet-produced particles which interact with the vapor flow. A complete model has been set up in order to calculate the jet breakup length and the generated particles’ diameter under each specific situation characterized by initial conditions. This model mainly relies upon results from boundary layer theory and linear instability calculations. The full model’s results are compared to existing experiences in this field and a final correlation of the results is established. A good agreement is obtained on the jet breakup length, however the predicted particle diameter tends to be too large. This last result could be explained by a secondary breakup of the particles in water and by a large uncertainty in the vapor flow.     

Deformation analysis of fuel pins within the wire-wrap assembly of an LMFBR

Deformation of fuel pins within the wire-wrap fuel assembly of a fast breeder reactor is analyzed by two computational codes, the subchannel deformation analysis code SHADOW and the thermal-hydraulic analysis code DIANA. Coupling these codes makes it possible to analyze percisely the mechanical interactions between all fuel pins in an assembly, and the deviation of coolant temperature distribution in deformed flow channels from the nominal distribution.In this paper, particular attention is paid to the effect on fuel pin deformation of the following factors: dimensional changes in the fuel assembly components, displacement of wrapper tube walls and changes in the radial power gradients.     

Fatigue strength of dissimilar welded joints for the main vessel of an LMFBR

The employment of welded joints composed of dissimilar metals is one simple and inexpensive way to connect a main vessel made of austenitic stainless steel and a roof slab constructed of ferritic steel in the design of liquid metal fast reactors. Since dissimilar-metal welded joints have not been used for such large structures so far in Japan, the structural integrity of this type of joint should be carefully examined for such a design option to be selected. Here various kinds of tests were conducted for eleven types of welded joints of 50 mm thickness to obtain this fundamental strength characteristics. Type 304 stainless steel was used as one of the parent metals in all the joints. They differ from each other in regard to the type of ferritic steel, welding metal and welding procedure. Low-cycle fatigue tests were conducted for round-bar specimens made from these welded joints at room temperature. Fatigue crack-propagation tests were also conducted for some of the joints. Tests after manufacturing a large-scale shell model were also conducted. The results of these tests demonstrated that the present manufacturing technique can, produce welded joints of high quality and reliability. A trial calculation for actual design conditions showed the existence of large margins against fatigue failure or fatigue crack-propagation of a significant amount.     

Experimental study on the effect of fission product gas release into blockage wake region using a simulated LMFBR fuel subassembly

The objective of this study is to evaluate temperature rise due to gas release in the wake region of LMFBR fuel subassemblies. The experiments were conducted in two sets of grid-spacer-type 37-pin bundles simulating LMFBR fuel subassemblies. In Test section 37GC, the central 24 subchannels were blocked by a stainless steel plate and in Test section 37GE one-half edge part (39 subchannels) of the total flow area was blocked by the same material. The experimental results were compared with data obtained in similar tests using a spacer wire-type pin bundle, designated 37WC. The temperature rises in 37GE and 37WC were nearly identical in value and effect of gas release rate. The marked agreement seems to imply that there is a limit in the content of released gas in the wake region. On the other hand, the temperature rise behind the central blockage in the grid-type bundle, where gas might easily flow out to the core flow region, was far less than in the other geometries.     

Application of an eddy-current type flowmeter to void detection at the LMFBR core exit

To apply an eddy-current type flowmeter to void (bubble) detection, out-of-pile test has been carried out in a sodium loop, and the void detection technique is applied to the subassembly outlet flowmeters of JOYO. The experimental apparatus used 7 simulated fuel subassemblies under 500°C sodium conditions with a volumetric void fraction of up to 2%. It was found that the phase of the void signal (vector) is different from that of the flow signal (vector). Therefore, the signal-to-noise ratio of void detection was highly improved by choosing the optimum phase angle for synchronous detection. At the optimum phase, the flow-induced fluctuation signal (background) is minimum and the void signal nearly maximum. In addition to the improved void-detection technique, basic information about amplitude probability density functions and power spectral densities of void signals are presented. Using the flowmeter installed in the core of JOYO, void-detecting characteristics have been studied.