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IN THIS VOLUME


Thermal Hydraulics

2010;():1-14. doi:10.1115/ICONE18-29008.

In-Vessel Retention (IVR) of core melt is a key severe accident management strategy adopted by operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000 etc. External Reactor Vessel Cooling (ERVC), which involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris relocated to the vessel low head, is a novel severe accident management for IVR analysis. In present study, IVR analysis code in severe accident (IVRASA) has been proposed to evaluate the safety margin of IVR in AP600 with anticipative depressurization and reactor cavity flooding in severe accident. For, IVRASA, the point estimate procedure has been developed for modeling the steady-state endpoint of core melt configurations. Furthermore, IVRASA was developed in a more general fashion so that it is applicable to compute various molten configurations such as UCSB FInal Bounding State (FIBS) etc. The results by IVRASA were consistent with those of the UCSB and INEEL. Benchmark calculations of UCSB-assumed FIBS indicate the applicability and accuracy of IVRASA and it could be applied to predict the thermal response of various molten configurations.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():15-26. doi:10.1115/ICONE18-29009.

Longitudinal Vortex (LV) is produced by Longitudinal Vortex Generators (LVGs) with high heat transfer efficiency and acceptable pressure loss. Due to the relative long influence distance and simple structure, LVGs may be used in narrow channels with flat surface under high temperature and high pressure water medium, in this paper, the critical heat flux (CHF) is one of most important focus. The test channel has the size of 600 mm (length) × 40 mm (width) × 3 mm (height), was used to research the CHF characteristic of CHF affected by LVGs. The test channel is visual in three sides and remains one side for power supply. The LVGs used in the experiments are 14 mm (length) × 2.2 mm (width) × 1.8 mm (height) in dimensions, and periodically mounted on the inner wall of the steel plate. The parameters that are varied during the experiments as follows, system pressure from 0.43 to 0.85 MPa, inlet mass flow flux from 40.2 to 745.7 kg·m−2 ·s−1 , inlet subcooling from 46.8 to 104.2 °C, exit quality from 0.183 to 0.997, surface heat flux from 0.294 to 2.316 MW·m−2 . The experiments show that the CHF is improved by 24.3% while the total pressure drop through the test section is improved by 62.9%. The bubble growth and its evolutionary process in narrow rectangular channel with LVGs have been obtained during a short term when the CHF occurs, and it is found that the bubbles have been affected intensely by LV. Based on these experiment data, the growth and aggregation of bubbles have been depressed by LV, the mass, momentum and energy exchange between cold and hot areas in the test section have been strengthened. As a result, the heat transfer enhancement by LV can be explained by the destruction of thermal boundary layer.

Commentary by Dr. Valentin Fuster
2010;():27-36. doi:10.1115/ICONE18-29017.

MAAP5 (Modular Accident Analysis Program Rev. 5.0.0), developed by Fauske & Associates, Inc.’s (FAI) based on the MAAP4 code, is a severe accident analysis code. It is a computer program capable of simulating the response and mitigation actions of light water reactor nuclear power plants (NPPs), including advanced boiling water reactor (ABWR) during severe accident. A specific loss of all core cooling accident sequence, LCLP-PF-R-N, based on Final Safety Analysis Report (FSAR) of Lungmen (ABWR) NPP, was selected as a based case and simulated by the MAAP5 and MAAP4 codes. The MAAP5 and MAAP4 parameter files for Lungmen NPP were established based on Lungmen NPP design data and the MAAP5 and MAAP4 users’ guides. The main severe accident phenomena and the fission product release fractions associated with the LCLP-PF-R-N sequence were simulated. The purpose of this paper is to compare the analysis results of LCLP-PF-R-N sequence calculated by MAAP5 and MAAP4 codes. The two codes give similar results for important phenomena during the accidents, including core uncovery, core support plate failure, debris relocation to the lower plenum, vessel failure, passive flooder opens, containment overpressure protection system (COPS) activation, noble gases and volatile species (like CsI) release to environment, except for the amount of hydrogen production in core. MAAP5 predicts a greater amount of hydrogen production in core than that of MAAP4. This is because MAAP4 predicts earlier reactor pressure vessel (RPV) depressurization than that of MAAP5. That results in earlier steam exhaustion and oxidation reaction termination in core than those of MAAP5. This paper successfully demonstrates the severe accident of Lungmen NPP, and analysis results can provide useful information for the MAAP5 and MAAP4 users.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():37-45. doi:10.1115/ICONE18-29024.

An analytic investigation of the thermal exchanges in channels is conducted with the prospect of building a simple method to determine the Nusselt number in steady, laminar or turbulent and monodimensional flow through rectangular and annular spaces with any ratio of constant and uniform heat rate. The study of the laminar case leads to explicit laws for the Nusselt number while the turbulent case is solved using a Reichardt turbulent viscosity model resulting in easy to solve one-dimensional ordinary differential equation system. This differential equation system is solved using a Matlab based boundary value problems solver (bvp4c). A wide range of Reynolds, Prandtl and radius ratio is explored with the prospect of building correlation laws allowing the computing of Nusselt numbers for any radius ratio. Those correlations are in good agreement with the results obtained by W.M. Kays and E.Y. Leung in 1963 [1]. They conduced a similar analysis but with an experimental basis, they explored a greater range of Prandtl but only a few discreet radius ratio. The correlations are also compared with a CFD analysis made on a case extracted from the Réacteur Jules Horowitz.

Commentary by Dr. Valentin Fuster
2010;():47-56. doi:10.1115/ICONE18-29034.

In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally-insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modelled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the insulation material during reverse flow. This will certainly further improve the coolability of the core. The spacer grids were modelled as a strainer, which completely retains all the insulation material reaching the uppermost spacer level. There, the accumulation of the insulation material gives rise to the formation of a compressible fibrous cake, the permeability of which to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity. Before the switch over of the ECC injection from the flooding mode to the sump mode, the coolant circulates in an inner convection loop in the core extending from the lower plenum to the upper plenum. The CFD simulations have shown that after starting the sump mode, the ECC water injected through the hot legs flows down into the core at so-called “breakthrough channels” located at the outer core region where the downward leg of the convection roll had established. The hotter, lighter coolant rises in the centre of the core. As a consequence, the insulation material is preferably deposited at the uppermost spacer grids positioned in the breakthrough zones. This means that the fibers are not uniformly deposited over the core cross section. When the inner recirculation stops later in the transient, insulation material can also be collected in other regions of the core. Nevertheless, with a total of 2.7 kg fiber material deposited at the uppermost spacer level, the pressure drop over the fiber cake is not higher than 8 kPa and all the ECC water could still enter the core.

Commentary by Dr. Valentin Fuster
2010;():57-65. doi:10.1115/ICONE18-29044.

This study was performed at the Institute for Neutron Physics and Reactor Technology at Karlsruhe Institute of Technology and is addressed to establish the combined usage of the best-estimate code TRACE and the uncertainty and sensitivity analysis tool SUSA, for safety related investigations of current and future nuclear energy systems. In the frame of this paper the applicability to supercritical water related investigations is covered. Several Nusselt correlation for the heat transfer to supercritical water are available and have been evaluated in previous investigations but not one of them gave satisfying results. Hence, the consideration of uncertainty and sensitivity measures applied to the topic of heat transfer seems to be an appropriate way. In a first step a post-test analysis of an experiment was conducted. Results showed that with the help of uncertainty and sensitivity methods parameters which affect the results most could be identified. The most important parameter was of course the Nusselt correlation. In addition to the identification of important parameters, the experimental results were enveloped by the calculated results. That means, in the sense of safety related evaluation of designs for reactors operated with supercritical water, that key parameters (cladding temperature) can be calculated with a certain confidence.

Commentary by Dr. Valentin Fuster
2010;():67-74. doi:10.1115/ICONE18-29046.

In this paper, a meshless local Petrov-Galerkin (MLPG) method is given to obtain the numerical solution of the coupled equations in velocity and magnetic field for the steady magnetohydrodynamic (MHD) flow through a straight duct of rectangular section. Mehdi Dehghan has applied MLPG method to solve the MHD flow control equations at Hartmman numbers less than 40. The integration subdomain is properly adjusted to carry out the computations for Hartmman numbers from 5 to 400. Numerical results show that the MLPG method with adjusted integration subdomains can compute MHD problems not only at low values but also at moderate values of the Hartmann number with good accuracy and convergence.

Commentary by Dr. Valentin Fuster
2010;():75-80. doi:10.1115/ICONE18-29050.

Axial power distribution is one of the parameters that influence the occurrence of the dryout in nuclear fuel assemblies. Experimental data indicate that this influence is quite substantial, ranging from few to above ten percent of the total power. Thus accurate prediction of the dryout power for various power distributions has important implications on the economy and safety of nuclear power plants. The difficulty with capturing the influence of that parameter stems from the fact that during reactor operation practically unlimited number of power shapes can occur. This fact makes it very difficult to investigate the effect experimentally, and an analytical approach is needed. Various methods have been proposed in the past to capture the effect of non-uniform power distribution on dryout. These approaches can be divided into several categories, where the two main ones are as follows: (a) methods based on introduction of a shape factor, which is calculated from the known shape of the power distribution; (b) methods using certain integral parameters, such as the boiling length and the annular flow length, which are expressed as functions of axial power distribution. In the present approach a simplified annular flow model is used, in which the dryout occurrence is based on the prediction of the disappearance of the liquid film. The dependence of the dryout power on the axial power shape is obtained in a general analytical form. Based on this analytical solution, a new set of terms that govern the dryout power in channels with various axial power distributions is proposed.

Commentary by Dr. Valentin Fuster
2010;():81-99. doi:10.1115/ICONE18-29053.

This documentation summarized initial events assessment of auxiliary building as a result of internal flooding based on Chinese 1000MWe PWR. Detailed screening procedure of plant areas and quantitative frequency of each identified initial event are presented. In order to implement the analysis, the whole auxiliary building is divided into several areas, depending on elevation, isolation and function. Areas that are essential to plant operation and that house safe shutdown equipment are screened through a two-step process to determine the susceptibility to an internal flood. After this process, areas that are not eliminated are where flooding scenarios may occur. All possible flooding scenarios are reviewed to determine if they could result in any of the initiating events identified in the internal initiating events analysis which belongs to Level 1 PSA. For those flooding scenarios that do not result in an existing initiating event, a new initiating event is developed. Eventually, the frequency of each initiating event is achieved by identifying each flooding source and its failure mode, and correlating this to failure data. It must be paid attention to that the identification of internal flood hazards is plant specific since the routing of piping systems varies substantially from plant to plant. Therefore, the conclusion obtained from this issue may be not applicable for other nuclear power plant.

Topics: Floods
Commentary by Dr. Valentin Fuster
2010;():101-105. doi:10.1115/ICONE18-29059.

Heat transfer is analyzed from a different view in mixed convection in this paper. A concept, namely averaged heat transfer resistance coefficient, is used to describe heat transfer performance. For local position, heat transfer defined by generalized Fourier law is determined by fluid conductance and turbulence heat transfer. On the other hand, heat resistance over the cross section is the integer of the local resistance, where the weight, a function of spatial position, can be expressed by product of local heat transfer and temperature. To enhance heat transfer, it is crucial to reduce the heat resistance where the weight is big, namely near the wall. Heat transfer performance under different buoyancy effect is analyzed by the new. The results show that flow structure and heat transfer are closely connected by a straightforward expression. Heat transfer mechanism of enhancement and deterioration under different stages can be perfectly explained, which can predict heat transfer qualitatively.

Commentary by Dr. Valentin Fuster
2010;():107-115. doi:10.1115/ICONE18-29061.

This study studied the performance of perforated orifices by a new method, namely field synergy theory (FST). Pressure loss was obtained theoretically by connecting flow field and pressure loss, and pointed out the direction to optimization. To validate the theory, pressure loss was measured experimentally under different orifice structures by changing the parameters of porosity, thickness, hole diameter, hole distribution, and etc. Theoretical results showed that pressure loss was determined by the synergy of velocity and velocity gradient over the flow field passing the orifice. Pressure loss increased as the synergy being stronger. To decrease pressure loss, it was the most effective to decrease the synergy with the biggest weight. The comprehensive performance of a perforated orifice is better than a standard one. All the structure parameters should be optimized to improve the performance of a perforated orifice.

Commentary by Dr. Valentin Fuster
2010;():117-122. doi:10.1115/ICONE18-29069.

This paper presents the results of the TOSQAN tests performed in the field of the TOSQAN sump program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN). This work is devoted to study the heat and mass transfers induced by the water sump towards the containment atmosphere, for typical accidental thermal hydraulic conditions in PWR.

Topics: Containment
Commentary by Dr. Valentin Fuster
2010;():123-131. doi:10.1115/ICONE18-29092.

Air-water CCFL (countercurrent flow limitation) tests using the 1/5th scale rectangular channel and 1/15th scale circular tube simulating a PWR hot leg and both air-water and steam-water CCFL tests using the 1/3rd scale rectangular channel were previously carried out at Kobe University and Forschungszentrum Dresden-Rossendorf (FZD), respectively. In this paper, numerical calculations for the air-water CCFL tests at FZD using FLUENT6.3.26 are presented and compared with the experimental data at Kobe University and FZD. In the calculations, the VOF (volume of fluid) model or two-fluid (2F) model was used. Major results were as follows: (1) the calculated CCFL characteristics using the 2F model for the FZD tests agreed well with the 1/15th scale circular tube data obtained at Kobe University and the calculated results for full-scale PWR conditions, which supported the validity of the 1/3rd scale rectangular channel to simulate CCFL in circular tubes; and (2) comparison with the FZD data showed that the calculations using the 2F and VOF models overestimated the falling water flow rates.

Commentary by Dr. Valentin Fuster
2010;():133-139. doi:10.1115/ICONE18-29114.

The density wave oscillation (DWO) was investigated with parallel rectangular channel test sections, which have 2mm*25mm flow cross section. Test parameters are pressure 1MPa–10MPa, mass velocity 200kg/m2 s–800kg/m2 , and inlet subcooling 10°C–50°C. The experimental data show that heat flux rises with high mass velocity, inlet subcooling or system pressure at the stable/unstable boundary. In them, increasing mass velocity can greatly improve stability of this two parallel channel system. Period of the oscillation becomes shorter while mass velocity rises, but when inlet subcooling is increased, it becomes longer. The comparison between data from rectangular channels and round tubes indicates they have the same trend based on the dimensionless phase change number and subcooling number.

Commentary by Dr. Valentin Fuster
2010;():141-146. doi:10.1115/ICONE18-29125.

Spherical fuel elements are distributed randomly in the pebble bed reactor core and helium flow through the pebble bed to remove nuclear reaction heat. Pebble bed reactor core is usually treated as a uniform porous media flow in thermal hydraulic research. However the porosity distribution is nonuniform and the porosity near the wall increase sharply. A new random model is developed in this paper to investigate thermal hydraulic characteristics of pebble bed reactor core. Porosity assumption is based on porosity measurement of other research. Porosity simulation is divided into three parts according to the distance from wall. In the center of core, porosity is assumed to obey normal distribution, where average porosity is from the experimental relation based on statistical results. The mean and standard deviation of porosity distribution near the wall will increase because of the wall effect, where the distance from the wall is about three times of fuel ball’s diameter. The third part is zone from three times to five times of ball’s diameter departed from the wall. The wall effect of this zone is between center and the wall zone. Based on above assumption, a random porosity simulation is completed to apply in this research. COMSOL Multiphysics 3.5a software is used in this research. COMSOL Multiphysics are a calculation platform using proven Finite Elements Methods (FEM). In this research, Brinkman equation for porous media flow is applied in the simulation. Non-thermal Balance model is used in local heat transfer research between gas and pebble bed. A geometry model is built to simulate HTR-10. Temperature profile of variant porosity is gained from stationary analysis and comparison with uniform porosity is also discussed in the paper. For transient analysis, four cases simulation is carried out in the research. Case 1 and 2 simulate heat transfer phenomena with forced cooling system and with passive cooling system after reactor shut down. Way-Wigner-curve is used in Case 1 and Case 2 to simulate decay heat in the calculation. Case 3 and Case 4 simulate ATWS phenomena with natural convection and without natural convection system when blower is trip off in normal operation. Simulation results also are compared with some ATWS experiments and some discussion is done in the paper. From the results, it can be seen that random porosity will affect temperature distribution near the wall and make outlet temperature non-uniform greatly. The maximum temperature of variant porosity is much greater than the maximum temperature of uniform porosity at the same condition. Transient analyses of variant model show that passive cooling system can remove residual heat even in accident conditions when the blower trip off whether reactor shut down or not and the analyses results correspond substantially with experimental results. In general, variant porosity should be considered in the thermal hydraulic research of pebble bed reactor core. Variant porosity model can provide good prediction of heat transfer phenomena than uniform porosity model. Especially it can explain some transient analysis results.

Topics: Simulation , Porosity
Commentary by Dr. Valentin Fuster
2010;():147-152. doi:10.1115/ICONE18-29137.

The OTSG (Once-Through Steam Generator) is usually used in the integral nuclear power equipment which requires smaller size and better effect of heat transfer. The OTSG with double-side heat transfer component is presented in this paper. The heat transfer component is composed of straight tube outside and helix tube inside. In the both sides of the helix tube, the flow is spirally, therefore, the heat transfer is enhanced. The smaller the pitch, the stronger the spirally flow, the effect of heat transfer is better, but the flow resistance is raised. Especially the increased flow resistance in the secondary side brings a great influence to the pump. The heat transfer region of the secondary fluid are divided into three regions: sub-cooled region, boiling region, and superheated region, the effects of heat transfer induced by the spirally flow vary in different regions. Thus, there is an optimization problem which is to find an optimization pitch of the inner helix tube with the best effect of heat transfer and the minimum flow resistance. Based on analyzing the effects of the pitch on heat transfer enhancement and flow resistance, the pitch is optimized by the constrained nonlinear optimization method.

Topics: Optimization
Commentary by Dr. Valentin Fuster
2010;():153-162. doi:10.1115/ICONE18-29148.

Heat transfer characteristic of low frequency pulsating turbulent water flow in a vertical circular pipe which is heated at uniform heat flux, are experimentally studied under different conditions of Reynolds number, pulsation frequency and relative amplitude. The experiments are performed with the Reynolds number range of 3000 to 20000, pulsation frequency range of 0.033 to 0.1 Hz, and the relative amplitude range of 0.1 to 0.8. This pulsating flow situation is used to simulate the phenomenon happened in the ship power system which is induced by ocean conditions. The effects of pulsation on heat transfer characteristics are presented in terms of relative local and mean Nusselt numbers defined as the ratio of the local and mean Nusselt numbers for pulsation flow to that of the ordinary steady turbulent flow with the same time-averaged Reynolds number Reta . The results show that the relative local Nusselt number is strongly affected by Reynolds number, pulsation frequency and relative amplitude. The phenomena that the Nusselt number would increase or decrease with the increase of the Reynolds number are both observed and the variation is more notable in the entrance region than that in the fully developed region. The relative mean Nusselt number decreases initially as the Reta increases, and then recovers gradually, but finally it has the tendency to decrease again. With the increase of pulsation relative amplitude, the relative mean Nusselt number increases at first and then decreases. And for the Reynolds number range of 3176 to 6670, heat transfer enhancement is observed as the pulsation frequency raises, but complete contrary phenomena appears at Reynolds number range of 11904 to 15844. The obtained heat transfer results are analyzed and seem to be qualitatively in accordance with previous investigations.

Commentary by Dr. Valentin Fuster
2010;():163-168. doi:10.1115/ICONE18-29151.

Cores loaded with a mixture of fuel types are known to reduce stability margins. Mixed fuel cores have become more common as utilities change fuel suppliers, or when fuel vendors upgrade their fuel designs to take advantage of improved thermal and mechanical margins. This paper studies some of the physical processes that reduce the stability of mixed cores. A number of runs have been performed using the LAPUR6 stability code to evaluate the effect on mixed cores on the stability of a typical BWR. To this end, two fuel types have been set up with two different single-phase to two-phase pressure drop ratios by artificially adjusting the spacer and inlet orifice friction coefficients. The flow and pressure drop characteristics of both fuels have been matched at full flow, full power conditions. All manufacturers match the pressure drop of new fuels so that the flow distributions among the new and old fuel elements operating at the same power are approximately constant. The critical power ratio and thermo-mechanical criteria are typically limiting at full power; therefore matching the flow performance at full power maximizes the margin to these criteria. Stability is of concern at low flows, especially at natural circulation, where the thermal-hydraulic conditions are significantly different from full flow and power. Our simulations show that even if two fuel elements are perfectly matched at full flow, the axial void fraction distribution changes significantly when the flow is reduced to natural circulation conditions and the two fuel elements are not fully thermal-hydraulically compatible at the reduced flows. Basically, the two fuel types set up two separate natural circulation lines, and one of the fuel types essentially starves the other from flow. Since stability has such a strong dependence with channel flow, the reactor stability is controlled by the fuel type that has the smaller flow at natural circulation. A counterintuitive result of this study shows that, in general, loading a more stable fuel type into a mixed core has the opposite effect, and the stability margin of that mixed core is lower until the new, more stable fuel becomes dominant. Because of the burnable Gadolinium in most modern BWR fuels, the highest reactivity fuel elements are the once-burned. Loading a more stable fuel type starves the flow of the high-reactivity older fuel, reducing the stability margin.

Commentary by Dr. Valentin Fuster
2010;():169-171. doi:10.1115/ICONE18-29153.

Most fuel modeling computer codes of a decade ago did not explicitly account for fuel thermal conductivity degradation with burnup. The overestimation of thermal conductivity is compensated by other code adjustments. Following the approval of the legacy fuel thermal codes, the quantity, quality and comprehensiveness of fuel measurement data have increased dramatically. Recently, the US NRC has informally requested the nuclear industry review the thermal performance codes with the consideration of thermal conductivity degradation. This paper compares the legacy thermal conductivity models with the most recent available models that are published publicly and may have been incorporated into present generation fuel performance codes. Also included in this paper is one method which AREVA NP Inc. developed to evaluate the thermal conductivity degradation effect. The effect of thermal conductivity degradation on the PWR licensing limits, such as centerline fuel melt and strain, is demonstrated.

Commentary by Dr. Valentin Fuster
2010;():173-178. doi:10.1115/ICONE18-29172.

The present studied Pebble-Bed Reactor is a light-water cooled reactor that consists of millions of Micro-Fuel Elements, and the TRISO-coated fuel particles (MFE) fill the fuel assembly disorderly and form a porous media with internal heat source. Papers on porous media continue to be published at the rate of about 150 per year and the domain of application is wide spread, ranging from chemical particle beds, mass separator units, debris beds, soil investigations, heat pipes and fluidized beds etc. In this paper, investigation is performed on the press drop under conditions of both single-phase and two-phase flow through porous media. Large number of relations are studied and the relational expressions, which generalize the available data of experiments, are suggested for pressure drop calculation in a pebble bed of spheres at random distribution. Finally, the relational expressions are applied to analyze the flow characteristics of the Pebble-Bed Reactor, such as the influence of pressure on two phase friction factor in the core etc.

Commentary by Dr. Valentin Fuster
2010;():179-186. doi:10.1115/ICONE18-29179.

Piping systems of nuclear power plants include connections of branches conveying fluids at different temperatures. Thermal-hydraulic fluctuations arising from the turbulent mixing of the flows can affect the inner wall of the pipes and lead to fatigue damage. In order to assess the high-cycle thermal fatigue damages risks of the NPP mixing zones, knowledge of the temperature fluctuations and heat transfer from fluid to structure is necessary. In order to have a better knowledge of the thermal loadings in the several kinds of mixing zones of a NPP, a multi-annual R&D program has been initiated by EDF, AREVA and CEA. The experimental program uses two kinds of representative small-scale mock-ups which are called “Skin of Fluid” and “Stainless Steel FATHERINO”, for different mixing configurations. “Skin of Fluid” mock-ups are very thin ones, made of brass material, allowing visualization with an infrared camera of the fluid temperature field in the mixing zone, with minimized thermal attenuation. As a consequence, “Skin of Fluid” mock-ups allow detecting areas of interest in the mixing zone. “Stainless Steel FATHERINO” mock-ups are then used to evaluate quantitatively the thermal fluctuations and the heat transfer coefficient in the mixing zone. “Stainless Steel FATHERINO” mock-ups are made of 304L stainless steel and are 9.5mm thick. The mock-ups are instrumented with specific sensors named “Coefh”. These sensors allow measuring simultaneously the fluid and structure temperature time-histories and thus determining the heat transfer coefficient. This paper describes the experimental program and the approach used to evaluate the thermal load in the mixing zones. As an example, the application of this approach to an equal T-junction is presented hereafter.

Commentary by Dr. Valentin Fuster
2010;():187-191. doi:10.1115/ICONE18-29187.

This paper provides a discussion of the model development status and verifications for the components used in the Reactor Thermal-Hydraulic model for the Full-Scale Simulator (FSS) of the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). Heat transfer network and thermal-hydraulic network are adopted to simulate the thermal-hydraulic process of reactor of HTR-PM. Due to the First of a Kind Engineering nature and lack of reference plant data, model verification will mainly focus on benchmarking the model configurations against test cases performed by HTR-PM design code THERMIX in future.

Commentary by Dr. Valentin Fuster
2010;():193-204. doi:10.1115/ICONE18-29200.

Two fluid model integrating a set of closure relationships (such as inter-phase heat transfer model, inter-phase mass transfer model, inter-phase momentum transfer model, mean bubble diameter model, bubble departure diameter model, bubble departure frequency model, onset of nucleate boiling model, wall heat flux partition model) are applied to solve the local flow and heat transfer of subcooled flow boiling under low pressure, and local flow parameters such as volume fraction, liquid velocity, vapor phase etc. are obtained using developed code. The subcooled flow boiling results predicted in this paper are compared to the subcooled flow boiling experimental results of TH Lee etc. in annular channel with inner tube heated uniformly and adiabatic outer tube, and they agree well. The code can also be used to predict the multiphase flow field in power reactor core when the subcooled flow boiling take place and the critical heat flux of Departure Nucleate Boiling (DNB) in reactor core and it is meaningful for reactor core design.

Commentary by Dr. Valentin Fuster
2010;():205-214. doi:10.1115/ICONE18-29201.

In the design of some passive PWRs, Reactor Coolant Pump (RCP) is welded directly to the Steam Generator (SG) channel head. This design cancels the support of RCP and simplifies the layout of Reactor coolant system. What’s more, this design also do good to the mitigation of SBLOCA. But this design makes the flow field in the SG channel head and RCP inlet complex and there may be vortex in this flow field for which reason the SG outlet resistance will increase and affect the long-term steady operation. For this issue, some company made tests on it. But the cost of test is high and the applicability of the test result is limited. If the parameters or components size changed a little, the test result will be no longer applicable. To solve this problem, this article considers using 3-D CFD flow field analysis software to analyse the SG and RCP coupled flow field. Through steps of 3D model establishing – meshing in Gambit – analyzing in Fluent, this paper obtains the flow filed condition of SG-RCP coupled part during normal operation and so as to support plant design.

Commentary by Dr. Valentin Fuster
2010;():215-222. doi:10.1115/ICONE18-29208.

To investigate the effect of mixing-vane shape, heat flux at departure from nucleate boiling (DNB) and pressure loss were measured. Computational fluid dynamics (CFD) was utilized to discuss the flow control. The pressure loss and the DNB tests were performed in a water and a Freon loops, respectively. Two mixing-vanes were designed to have same projection area but different inclination. The rod-bundle was 5 by 5 and 17 by 17 respectively at the water and Freon tests. The experimental results showed that the slightly inclined mixing-vane produced the same DNB heat flux as the deeply inclined mixing-vane and did smaller pressure loss than it. Pressure loss of the two mixing-vane grids was different in spite of the same projection area. The result of CFD showed a swirl flow decaying along the main stream in the axial direction. The swirl was stronger in the deeply inclined mixing-vane, however it decayed faster whereas one maintained long in the slightly inclined mixing-vane. This result suggested that the deep inclination caused a steep change in axial momentum to induce strong turbulence diffusion. This flow structure did not change the DNB heat flux because the two-phase discontinuity dominated the phenomena. This study provided a successful example of flow control in a mixing-vane grid.

Topics: Flow control
Commentary by Dr. Valentin Fuster
2010;():223-227. doi:10.1115/ICONE18-29210.

3-D single-phase flow field in 5×5 rod bundles with spacer grids was studied by numerical method. Using hybrid grids technique, SST k-ω model and SIMPLEC algorithm, the Reynolds averaged mass conservation and momentum conservation equations were solved, and the pressure and velocity field were obtained. The simulation results show that the spacer grids leads to intense lateral flow in rod bundles channel, and it submitted parabola distributing raw; axial velocities were distributed uniformly in the channel; drag coefficient decreased as inlet Reynolds number increased. Calculated results agreed with experiment data well, it shows that numerical methods employed in this paper is suitable to study flow field in 5×5 rod bundles.

Commentary by Dr. Valentin Fuster
2010;():229-234. doi:10.1115/ICONE18-29220.

In this paper, a meshfree point collocation method, with a upwinding scheme, is presented to obtain the numerical solution of the coupled equations in velocity and magnetic field for the fully developed magnetohydrodynamic (MHD) flow through a straight pipe of rectangular section with insulated walls. The moving least-square (MLS) approximation is employed to construct shape functions in conjunction with the framework of point collocation method. Computations have been carried out for different applied magnetic field orientations and different Hartmann numbers from 5 to 1,000,000. As the adaptive upwinding local support domain is introduced in the meshless collocation method, numerical results show that the method can compute MHD problems not only at low and moderate values but also at high values of the Hartmann number with high accuracy and good convergence.

Topics: Flow (Dynamics)
Commentary by Dr. Valentin Fuster
2010;():235-241. doi:10.1115/ICONE18-29228.

Experiments were performed on the double-tube bundle heat exchanger. The hot side of the heat exchanger consists of parallel flow channels of the inner tubes and the space between the shell and the double-tube bundle, and the cold side is the narrow annular flow channels between the inner and outer tubes. The test fluid in both hot and cold sides is de-ionized water. The heat-transfer coefficients were obtained from the experiment in the range of Reynolds number from 2000 to 38000. The detached coefficients method was used in the absence of direct measurement of the tube temperature. It is found that the ranges of laminar flow in both internal and external flow passages are wider than regular circular tube. The heat transfer coefficient of internal side is lower than those from Dittus-Boelter formula, while the heat transfer coefficient of external side in this experiment is higher than that for the corrugated pipe. The heat transfer coefficient of the narrow annular channel is significantly determined by its structure.

Commentary by Dr. Valentin Fuster
2010;():243-248. doi:10.1115/ICONE18-29239.

On the basis of best estimate thermal-hydraulic system code RELAP5, sub-channel code COBRA-1V, and commercial Computational Fluid Dynamics (CFD) code CFX, a thermalhydraulic multi-scale coupled code RECOX has been developed. The coupling strategy was designed to keep the integral structure of each code and minimize modifications of code source. Under the Parallel Virtual Machine (PVM) environment, an external control code has been developed to perform codes spawn, data exchange and mapping, time step coordination, etc. Two test cases including single phase blowdown and temperature fluctuation transient have been carried out to evaluate the coupling between codes. Compared with stand-alone simulations very good agreement was achieved. Then in order to demonstrate the coupled analysis capability of RECOX, an asymmetry transient in a simple two loops system which is similar to the nuclear power plant was simulated. The result is correct and reliable, although further verification of coupled code with related experiment is needed. Finally, some potential improvements of coupling and future work were presented.

Commentary by Dr. Valentin Fuster
2010;():249-255. doi:10.1115/ICONE18-29242.

A code has been developed with proper models for the thermal-hydraulic simulation of research reactors, and the accident of loss of offsite power of CARR (the China Advanced Research Reactor) has been analyzed. It is found that the transient can be departed into four parts with two peaks, and at last terminated to a natural circulation state. Furthermore, in order to investigate the effect of the height of the in core structures during this accident, several structure changes have been proposed and analyzed. It is demonstrated that the increase of the vertical height of the in core structures not always do help to decreasing the peak temperatures and increasing the natural circulation flow in the transient.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():257-262. doi:10.1115/ICONE18-29259.

This paper shows the experiments of the flow rate distribution measurements of air-water two-phase flow in vertical multi-branch carried out on the Gas-liquid two-phase flow fluid and flow rate distribution experimental platform, and discusses the laws and the Characteristics of two phase flow distribution in the multi-branch by changing the water flow rate or the air flow rate under the condition of fix the flow rate of the other phase, then analyzes the degree of the flow deviation of both two phases in branches by calculating the standard deviations of each working condition. The experiments show that the flow distributions of both two phases are very uneven. There was much air but little water flows into the branches which were close to the inlets of the distributor, but in the branches which were far away from the inlets of the distributor, the situation was opposite to the previous one. The types of the flow pattern in each branch in every working condition were obtained through the Hewitt-Robert flow pattern picture. As the flow rate of air increased, the annular pattern would spread from the branches which are close to the inlets of the distributor to the ones which are far away from the inlets of the distributor.

Commentary by Dr. Valentin Fuster
2010;():263-268. doi:10.1115/ICONE18-29278.

The thermal stratification is generic to un-isolable piping system in pressurized water reactor (PWR), which is not considered in original design. This effect can finally threaten the integrity of the piping system. The pressurizer surge line is affected by thermal stratification during reactor heat up and shutdown processes particularly. It’s important to investigate and mitigate thermal stratification in the surge line to meet safety requirements. In this paper, the thermal hydraulic analysis and the mechanical analysis using numerical simulation method were conducted to evaluate the impact of thermal stratification in the surge line during reactor heat up processes. The results give the three-dimensional temperature distribution and stress status of the surge line. It indicates that thermal stratification can cause stress concentration in the displacement restricts. The mitigation of thermal stratification was also studied. The numerical simulation of pressurizer surge line can put forward thermal stratification analysis.

Commentary by Dr. Valentin Fuster
2010;():269-273. doi:10.1115/ICONE18-29280.

Plate fuel assemblies will be widely applied in the future because of their simple and compact structure and excellent heat exchange capability. When the plate fuel assemblies are used in a ship reactor, the effect of the ship motion on the reactor core thermal hydrodynamic must be investigated. In this paper, aiming at the rectangular shape of coolant channels in the assemblies, the numerical simulation of the flow and heat exchange in a rectangular channel in horizontal translation is carried out. Using the CFD Software and UDF code, the simulation model of the channel in horizontal translation with and without acceleration along its width direction is built up, and the related characteristics is analyzed and discussed. The results and conclusions are applicable and useful for the new type ship PWR design.

Commentary by Dr. Valentin Fuster
2010;():275-280. doi:10.1115/ICONE18-29284.

According to the specific structure of a marine reactor pressurizer and its operating characteristics of frequent pressure and water level fluctuation, a new non-equilibrium four region mathematical and physical model is established for the first time to fulfill the aim of fast and precise simulation. The model includes the bubble rising, condensate dropping, wall condensation and heat transfer through outside wall while the heat and mass transfer at the liquid and gas interface is neglected. The initializing program is also introduced. The Simulink is adopted as the simulating tool and the corresponding Simulink module is established. The simulation proves to be very fast in a common microcomputer, which indicates that Simulink is efficient and is suitable for the fast simulation of marine reactor with the characteristics of frequent variations of power and corresponding physical and thermal parameters. A typical process with surge flow varying very acutely and complicatedly is adopted as an example. The variation of the pressure and the water level of the process by the new model agree well with the results of Relap5/Mod3.3, which proves the validity of the model and some simplification adopted. As one of the main parts of the fast simulation program under development for in the operating field use for a marine nuclear power plant, the achievement of the fast simulation of the pressurizer is of significant importance.

Topics: Simulation
Commentary by Dr. Valentin Fuster
2010;():281-290. doi:10.1115/ICONE18-29296.

This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: debris-bed coolability and metallic fuel freezing behavior. Related experimental database were reviewed to choose suitable experiments. To analyze the debris-bed coolability, the ACRR-D10 in-pile experiments were selected. SIMMER-III well simulated the heat transfer mechanisms including conduction, boiling and channeling observed in the experiment. Metallic fuel may freeze onto the stainless steel (cladding or wrapper tube) together with eutectic formation during core disruption in a metallic-fueled reactor. The CAFÉ-UT2 experiment carried out using pure UO2 melt to investigate such phenomena was selected for the experimental analysis. In spite of no eutectic formation model in the SIMMER-III code, the calculated fuel penetration behavior was in good agreement with the experimental data.

Commentary by Dr. Valentin Fuster
2010;():291-295. doi:10.1115/ICONE18-29298.

Based on the reactor physics and heat transfer parameter of rod-shaped fuel element, the steady state distribution and variation for rod-shaped assembly are analyzed and computed; using the virtual reality technology, 3D geometric entities of fuel element is constructed and divided into grid according to different condition; Through the normalized treatment for numerical results, the VR simulation environment for temperature field of fuel element is constructed. The research for field visualization has some reference value for core thermal design and operational safety analysis.

Commentary by Dr. Valentin Fuster
2010;():297-303. doi:10.1115/ICONE18-29303.

In this paper, a hydraulic model for Safety Injection System (SIS) of M310 reactor is extended. The model is checked and calibrated by test results under test conditions. Based on commissioning test criteria, the system’s maximum and minimum pressure drop coefficients are calibrated according to anti-extrapolation method. Considering modifications of various projects, analysis of 41 flow rate curves under different conditions has been performed using this model. These flow rate curves indicate the relationship between injection flow rates and primary circuit pressures. Accident analysis has taken these curves as input data. Also, the strategy for dealing with accident is established based on these curves. Results of accident analysis show that the design of SIS system can satisfy the safety requirements of M310 reactor. The sensitivity analysis of typical conditions illustrates that injection flow rate will increase as the primary circuit pressure decreases. With the same configuration, the injection flow rate during recirculation phase will be smaller than that during direct injection phase, which is mainly caused by the decrement of suction elevation and the increment of fluid temperature. When low head safety injection pump (LHSI PO) is boosting high head safety injection pump (HHSI PO), if the pressure is relatively high, the injection flow rate will not be improved apparently. If the pressure is relatively low, the boosting is necessary. These conclusions can be the basis for the later optimization design.

Commentary by Dr. Valentin Fuster
2010;():305-316. doi:10.1115/ICONE18-29306.

The measurement of two-phase flow parameters has never been an easy task in the experimental thermal-hydraulics and the need of such measurements in the SPES3 facility has led to investigation of different possibilities and evaluation methods to determine mass flows and energies. This paper deals with the theoretical prediction of the two-phase mass flow rate by the development of a mathematical model for a spool piece, consisting of a drag disk, a turbine and a void fraction detector. Data obtained by simulation of DBAs in the SPES3 facility, with the RELAP5 thermal-hydraulic code, have provided the reference conditions for defining the main thermal-hydraulic parameter ranges and selecting a set of instruments potentially suitable to measure and derive the required quantities. The governing equation and the instrumentation output are defined for each device. Three different turbine models (Aya, Rouhani and volumetric) have been studied to understand which one better adapts to two-phase flow conditions and to investigate the best instrument combination. The mathematical model has been tested versus the RELAP5 results with a reverse process where calculated variables, like void fraction, quality and slip ratio, are given as input to a specifically developed program to get back the mass flow rate. The analytical results, verified versus the DVI break transient, well agree with the RELAP5 mass flow rate. Specific tests on proper experimental loops are required to verify the analytical studies.

Commentary by Dr. Valentin Fuster
2010;():317-321. doi:10.1115/ICONE18-29308.

In this paper, a simultaneous visualization and measurement study have been carried out to investigate bubble nucleation frequency of water in micro-channel at various heat fluxes and mass fluxes. A single micro-channel with an identical rectangular cross-section having a hydraulic of 137 μm and a heating length of 30 mm was used in this experiment. It is shown that the frequency of bubble nucleation increased drastically with the increase of heat flux and was also strongly dependent on the mass flux. A dimensionless frequency of bubble nucleation was correlated in terms of the Boiling number. The predictions of bubble nucleation frequency in the microchannel are found in good agreement with experimental data with a MAE of 10.4%.

Commentary by Dr. Valentin Fuster
2010;():323-331. doi:10.1115/ICONE18-29319.

The Song, Mason and Ihm [1–4] equation of state with modification of Tao and Mason [5], originally derived for spherical and molecular fluids, is applied for fluid uranium tetrafluoride and thorium tetrafluoride based on the available experimental data. The equation of state based on statistical mechanical perturbation theory with the perturbation scheme of Weeks, Chandler, and Andresen [6]. The prediction of constants applied in the equation of state is based on the work of Boushehri et al. [7–8] using heat of vaporization and liquid density at the triple point. The vapor correction term bases on the experimental data. The results show that this equation of state agrees reasonably well with the available experimental data.

Commentary by Dr. Valentin Fuster
2010;():333-342. doi:10.1115/ICONE18-29320.

RD-14M experimental facility is a full vertical-scale representation of a CANDU heat transport system, that was used as a benchmark data generating facility within the frame of IAEA’s Technical Working Group on Advanced Technologies for HWRs. RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for an international standard problem exercise. The aim was the intercomparison and validation of computer codes for thermalhydraulics safety analyses, using both codes originating within the HWR technology and also different versions of RELAP5 code, the later being developed for transient simulation of light water reactor coolant systems during postulated accidents. A report was published by IAEA in 2004. The Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE), is also a LWR safety analysis code, developed jointly by AREVA_NP (reactors vendor), CEA (the French Atomic Energy Commission), EDF (the French electricity utility) and IRSN (the French Nuclear Safety Institute). The paper presents a model created for the RD-14M facility with CATHARE2, and the code application to the B9401 test. From the extensive series of test results available, several were selected to be compared to corresponding calculated evolutions. In some cases, our results are placed among those produced by the participants to the international standard problem with other codes.

Topics: Simulation
Commentary by Dr. Valentin Fuster
2010;():343-347. doi:10.1115/ICONE18-29328.

As a non-intrusive, whole-field temperature measurement technique, LIF (Laser Induced Fluorescence) has been used successfully to measure temperature fields. The performances of dyes are essential of the technique, especially the temperature sensitivity of the dyes. This work presents an analysis to provide a correct choice of temperature sensitive dyes combination (FL27 and RhB). The influences of temperature, excitation wavelength and pH on emission intensity and temperature sensitivity were analyzed. The results show that the temperature dependent tendency of FL27 changed from negative to positive as the excitation wavelength increased. The temperature sensitivity (4.0% per °C) of combination under 532nm laser is better than that of the wide used combination of RhB and Rh110 (2.0% per °C). The emission intensities of dyes are sensitive to pH value; however, the temperature dependence is unaffected.

Commentary by Dr. Valentin Fuster
2010;():349-354. doi:10.1115/ICONE18-29333.

A computational multi-fluid dynamics (CMFD) code that predicted the void distribution in sub-channel was coupled with sub-channel analysis code to predict departure from nucleate boiling (DNB). The main assumption was that the void fraction near heated wall was the dominant parameter in DNB. A sub-channel analysis code was used to calculate three dimensional distribution of sub-channel averaged values of mass flux, void fraction, density and quality. These were used as a boundary condition in the CMFD code to predict local void fraction in a subchannel. A bubble diffusion equation was used assuming the wall peak void distribution caused by turbulence. The present method was applied to the analysis of DNB tests. The coupled codes showed a reasonable profile of void fraction in a rod bundle and reproduced DNB heat flux at low void fraction. To investigate this analysis result, the local condition were compared with a DNB flow regime map. This suggested an approach to improve the predictability: the critical void fraction should be modified at low void fraction condition; the bubble diffusion model should be modified to handle the flow regime transition from the isolated nucleation type to churn turbulent flow type.

Topics: Fuel rods
Commentary by Dr. Valentin Fuster
2010;():355-368. doi:10.1115/ICONE18-29335.

The subcooled boiling heat transfer (HT) and the steady-state critical heat fluxes (CHFs) in a short SUS304-tube with twisted-tape insert are systematically measured for mass velocities (G = 4016 to 13850 kg/m2 s), inlet liquid temperatures (Tin = 285.82 to 363.96 K), outlet pressures (Pout = 764.76 to 889.02 kPa) and exponentially increasing heat input (Q = Q0 exp(t/τ), τ = 8.5 s) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304 test tube of inner diameter (d = 6 mm), heated length (L = 59.5 mm), effective length (Leff = 49.1 mm), L/d (= 9.92), Leff /d (= 8.18) and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.18 μm) is used in this work. The SUS304 twisted tape with twist ratio, y [= H/d = (pitch of 180° rotation)/d], of 3.39 is used. The relation between inner surface temperature and heat flux for the SUS304-tube with the twisted-tape insert are clarified from non-boiling to CHF. The subcooled boiling heat transfer for SUS304-tube with the twisted-tape insert is compared with our empty SUS304-tube data and the values calculated by our and other workers’ correlations for the subcooled boiling heat transfer. The influences of the twisted-tape insert and the swirl velocity on the subcooled boiling heat transfer and the CHFs are investigated into details and the widely and precisely predictable correlations of the subcooled boiling heat transfer and the CHFs for turbulent flow of water in the SUS304-tube with twisted-tape insert are given based on the experimental data. The correlations can describe the subcooled boiling heat transfer coefficients and the CHFs obtained in this work within −25 to +15% difference.

Commentary by Dr. Valentin Fuster
2010;():369-375. doi:10.1115/ICONE18-29341.

The narrow annular channel has been widely studied for its relatively larger heat transfer surface and structural compatibility. In this study, numerical studies have been performed on the 3D forced flow and heat transfer of water in concentric and eccentric annuli by using CFX codes. The gaps of concentric annuli range from 1.0mm to 4.0mm with the interval of 0.5mm. The eccentricity ratioes in eccentric annuli are 0.2, 0.3, 0.5 and 0.7. The radius ratioes of the eccentric annulus include 0.33, 0.5 and 0.66. The calculated results are compared with some experimental data and they agree well. The results show that the flow frictional resistance factor decreases with increasing the gap size. The impact of gap sizes on the flow frictional resistance factor decreases with increasing Reynolds number. The flow frictional resistance factor in the eccentric annuli is larger than that in concentric annuli. Furthermore, the effects of the eccentricity ratio and gap size on Nu number and the flow frictional resistance factor are also investigated.

Commentary by Dr. Valentin Fuster
2010;():377-382. doi:10.1115/ICONE18-29345.

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.

Commentary by Dr. Valentin Fuster
2010;():383-387. doi:10.1115/ICONE18-29351.

Two-phase natural circulation flow instability under rolling motion condition was studied experimentally and theoretically. Experimental data were analyzed with nonlinear time series analysis methods. The embedding dimension, correlation dimension and K2 entropy were determined based on phase space reconstruction theory and G-P method. The maximal Lyapunov exponent was calculated according to the methods of small data sets. The nonlinear features of the two phase flow instability under rolling motion were analyzed with the results of geometric invariants coupling with the experimental data. The results indicated that rolling motion strengthened the nonlinear characteristics of two phase flow instability. Some typical nonlinear phenomena such as period-doubling bifurcations and chaotic oscillations were found in different cases.

Topics: Motion , Time series
Commentary by Dr. Valentin Fuster
2010;():389-399. doi:10.1115/ICONE18-29356.

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():401-413. doi:10.1115/ICONE18-29386.

Under OECD PSB-VVER project, five experiments were performed in the PSB-VVER test facility. They are: 11% upper plenum break, natural circulation test, small break in cold leg, leak from primary circuit into secondary one and large break in cold leg. The experimental program purposes to cover the conditions that are highly relevant to the code validation related to PWR in general and to VVER-1000 safety assessments in particular. Main results of the OECD PSB-VVER project are presented in the paper.

Commentary by Dr. Valentin Fuster
2010;():415-423. doi:10.1115/ICONE18-29387.

The possibility to adequately simulate the flow circulation in the atmosphere of a nuclear power plant containment at accident conditions using a lumped-parameter code is investigated. An experiment on containment atmosphere mixing and stratification, which was performed in the TOSQAN experimental facility at IRSN in Saclay (France), was considered. During some phases of the experiment, steady states were achieved by keeping the boundary conditions constant. Two steady states during which natural convection was the dominant gas flow mechanism were simulated independently with the CFD code CFX4, whereas the entire transient was simulated with the lumped-parameter code ASTEC. The nodalisation of the lumped-parameter model was based on the flow pattern during the considered steady states, simulated with the CFD code. The lumped-parameter simulation succeeded in replicating the basic pattern of natural circulation in the vessel atmosphere during the two considered steady states, determined with the CFD code.

Commentary by Dr. Valentin Fuster
2010;():425-432. doi:10.1115/ICONE18-29394.

For the study of the Heterogeneous Inherent Boron Dilution transient in a Pressurized Water Reactor, a Small Break Loss Of Coolant Accident (SB-LOCA) is postulated. Natural Circulation (NC) may be interrupted and, under Reflux-Condenser (RC) conditions, the steam formed in the core condensates in the Steam Generator (SG) U-tubes: a boron-depleted slug may accumulate in the crossover leg and in the SG outlet chamber. If NC restarts as the Reactor Cooling System (RCS) is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. The mixing of the boron depleted slug with the borated water in the Cold Legs (CLs), downcomer and lower plenum after Restart of Natural Circulation (RNC) is quantified by means of Computational Fluid Dynamics (CFD) analyses. The CFD code STAR-CD is used to perform this analysis. Boundary conditions for this calculation — especially the boron-depleted slug size and the NC restart mass flow rate — are extrapolated from PKL experimental findings. The initial conditions are derived from an overall plant analysis performed with the CATHARE system code. Buoyancy effects, both in the cold leg and in the downcomer, are very significant phenomena for the evaluation of the slug transport and mixing: the hot (saturation temperature) boron-depleted water slug tends to accumulate in the upper parts of the cold legs and in the upper part of the downcomer (above the cold legs), before being pushed and dragged down. The boron concentration distribution at the core inlet during the transient, evaluated with STAR-CD, is compared with a critical value in order to check that boron concentration at the core inlet is always above the threshold necessary for the core to remain subcritical.

Commentary by Dr. Valentin Fuster
2010;():433-439. doi:10.1115/ICONE18-29400.

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

Commentary by Dr. Valentin Fuster
2010;():441-448. doi:10.1115/ICONE18-29411.

In the framework of accelerator driven sub-critical reactor systems (ADS), heavy liquid metals (HLM), in particular lead or lead bismuth eutectic (LBE), are considered as coolant for the reactor core and the spallation target due to their efficient heat removal properties and high production rate of neutrons. LBE-flows are characterized by excellent heat conductivity and exhibit a low molecular Prandtl number of the order 10−2 leading to distinct thermal and viscous boundary layers and scale separation in both the time and spatial domain. Since the analogy of turbulent heat and momentum transfer is employed in common turbulence models but is not valid in HLM flows, commercially available fluid dynamic code systems cannot predict heat transfer adequately for such flows. In order to provide validation data and heat transfer correlations, a series of three major experiments has been launched at the KArlsruhe Liquid metal LAboratory (KALLA) of the Karlsruhe Institute of Technology and will be presented in this overview.

Commentary by Dr. Valentin Fuster
2010;():449-454. doi:10.1115/ICONE18-29413.

Research and development programs of high safety significance have been going on for nuclear power plants of VVER-440/213 type to apply the in-vessel corium retention concept. The in-vessel retention (IVR) concept is based on the external reactor vessel cooling (ERVC) with the main objective to prove that the reactor pressure vessel (RPV) integrity can be preserved in accident sequences leading to core melt. One of the bases of programs was the SARNET project of European Commission, which focused on confirming the capability of the ASTEC code to simulate IVR, calculating thermal load caused by the corium. The ERVC concept is applied to the Paks nuclear power plant of VVER-440/213 type. For the experimental modelling of the ERVC, the CERES (Cooling Effectiveness on Reactor External Surface) facility was designed and constructed. The facility is a scaled down model of the cooling system intended to apply to the Paks NPP with 1:40 scaling ratio for the vessel external surface and 1:1 for the elevations giving the driving force for natural circulation. The heat load supplied to the model is provided by electric heaters. A large number of temperature, pressure, level, void and flow measurements are installed. A RELAP model of the CERES facility was developed and tested by pre-test results.

Commentary by Dr. Valentin Fuster
2010;():455-463. doi:10.1115/ICONE18-29414.

Several intermediate heat exchanger (IHX) modelling techniques were examined, in order to predict the outlet temperature of primary and secondary sodium at different operating conditions. In the present study, two different approaches namely the Finite Difference Method (FDM) with nodal heat balance and modified nodal heat balance schemes; and Finite Volume Method (FVM) using simple upwind, exponential extrapolation and QUICK schemes have been attempted.

Commentary by Dr. Valentin Fuster
2010;():465-470. doi:10.1115/ICONE18-29419.

Subcooled boiling in natural circulation has great influences on the flow and heat transfer characteristics of the medium in primary circuit loop of pressurized water reactor. Onset of nucleate boiling (ONB) is a key point in boiling heat transfer. According to analyze the existing models of ONB, the main factors that affect ONB are found out. The effects of main factors and the location of ONB are analyzed based on Yang Ruichang model in natural circulation, and found a model to calculate the location of ONB in natural circulation. A mathematical model is created using unascertained mathematics and compared with other models. The conclusion illustrates that the predicting model can describe the physical phenomena of subcooled boiling efficiently, and depicts the relation of certainty and uncertainty of influencing factors preferably by unascertained mathematics, which has particular methods to handle incomplete information, unsteady and inaccuracy numerical value, and variation quantity. The model reveals the real circumstance more accurately and plays an important role in security evaluation in nuclear power plant.

Commentary by Dr. Valentin Fuster
2010;():471-480. doi:10.1115/ICONE18-29449.

The common approach to safety in a nuclear power plant is to design the system to respond safely to a large postulated accident, the so-called design basis accident. Accidents more severe than the design basis accident (“severe accidents”) are assessed but the system is not designed to withstand them; they are considered too unlikely to require specific design actions. For the pressurised-water reactor (PWR), the design basis accident (DBA) is the Large Break Loss-of-Coolant Accident (LB-LOCA), in which it is assumed that one of the large inlet coolant pipes from the circulating pump to the reactor vessel is completely broken and moves apart to allow free discharge of the primary coolant from both broken ends. For this type of break total coolant loss will occur in 100s or less. Although the reactor is by this time sub-critical so that little power is produced from fission, a large amount of decay power exists and causes the fuel rod claddings to have a temperature in the region of 600–800 °C. This paper describes experimental investigations designed and performed in order to provide detailed information about the macroscopic behaviour of the steam-water flow occurring during the reflood phase following a PWR LB-LOCA. Specifically, a bottom-up rewetting process was studied, in which water droplets may be entrained in the vapour flow and contribute to cooling of the hot fuel pin before it is quenched. In these experiments the test section is initially preheated to temperatures up to 600 °C and then quenched by introducing water at the bottom of the tube at atmospheric pressure. During the course of this transient process axial temperature and heat flux profiles will be recorded, extending the existing databank of cases for code validation. Simultaneously, an axial viewing technique will be applied to observe the quench front, and any pre-cursory droplet production, occurring during these singletube reflood experiments. As part of the preliminary validation of this novel technique, a series of air-water vertical upflow conditions have been examined. The results of these preliminary studies provide detailed visualisation of typical entrainment processes likely to be encountered during single-tube reflood.

Commentary by Dr. Valentin Fuster
2010;():481-488. doi:10.1115/ICONE18-29452.

The paper presents the results of the OECD/NEA benchmark transient ‘Switching off one main circulation pump at nominal power’ analyzed as a boundary condition problem by the coupled system code ATHLET-BIPR-VVER. Of primary interest are the comparisons done for the local in-core parameters — assembly outlet coolant temperatures at 93 measured points and SPND powers at 7 layers of 64 fuel assemblies. Revealed are some sources of inaccuracy and methods are proposed to be decreased. An important step is done for the future performing of uncertainty and sensitivity analysis in the frame of the OECD/NEA activities.

Commentary by Dr. Valentin Fuster
2010;():489-497. doi:10.1115/ICONE18-29453.

The injection of a high pressure gas into a stagnant liquid pool is the characteristic phenomenon during the expansion phase of a hypothetical core disruptive accident in liquid metal cooled fast reactors. In order to investigate lot of mechanism involved in this phase of the accident’s evolution, an experimental campaign called SGI was performed in 1994 in Forschungszentrum Karlsruhe, now KIT. This campaign consists of nine experiments which have been dedicated to assess the effects of different pressure injection, the nozzle’s size and the presence of inner confinement in the formation of the rising bubble. Three of these experiments, which were focused on the pressure effects, have now been simulated with SIMMER III code and with FLUENT 6.3 code, a commercial CFD code. Both codes, despite their different features, have showed a good agreement with the experimental results. In particular, time trend evolutions of pressures and bubble volumes have been well reproduced by simulation. Furthermore, both codes agree on the shape of the bubble, even though they have evidenced same discrepancies with the experimental shape.

Topics: Simulation
Commentary by Dr. Valentin Fuster
2010;():499-506. doi:10.1115/ICONE18-29466.

The AP1000™ PWR reactor vessel upper plenum contains numerous control rod guide tubes and support columns. Below the upper plenum are the upper core plate and the top core region of the fuel assemblies. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions are performed, conducting local simulations for smaller sections of the domain can provide crucial and detailed physical aspects of the flow. These sub-domain models can also be used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. The study discussed in this paper focuses on the sections of the domain related to the control rod guide tubes. The commercial CFD code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier-Stokes equations for incompressible flow with a Realizable k-epsilon turbulence model, and to post-process the results. Two sub-domains are modeled and analyzed: (1) a 1/4 section of one control rod guide tube by itself and (2) a representative unit cell containing two sections of adjacent control rod guide tubes and one 1/4 section of a neighboring support column. For the 1/4 guide tube model (sub-domain 1), trimmed meshes of up to 16 million cells are generated to compute the flow and pressure fields in both complete and simplified (without chamfers and narrow gaps) models. Comparisons of the results lead to the conclusion that the simplified geometry model might be used when developing larger domain models in the future. The representative unit cell (sub-domain 2) is assumed to be positioned in the center of the upper plenum where the global lateral flow effects are minimal. At this position, the lateral flows are generated mainly by the flow as it exits the guide tubes. After flow enters the unit cell from the bottom, there are three potential locations for flow to leave the unit cell: (1) lower locations near the support column and the upper core plate, (2) side windows in the lower portion of the guide tubes, and (3) upper locations near the guide plates positioned inside the guide tubes. Both trimmed and polyhedral meshes are generated as part of the mesh sensitivity studies. Predicted flow rates at each of the outflow locations in conjunction with results from the mesh sensitivity studies provide guidance on (1) what geometry to preserve or remove, (2) what geometry can be simplified to reduce the required mesh, and (3) an estimate of the total mesh required to model the control rod guide tube and support column geometry in the much larger simulation of the entire upper plenum and top fuel domain.

Commentary by Dr. Valentin Fuster
2010;():507-515. doi:10.1115/ICONE18-29474.

The next generation nuclear plant (NGNP), whose development is supported by the U. S. Department of Energy, will be a very high temperature reactor (VHTR). The VHTR is a single-phase helium-cooled reactor that will provide helium at up to 800 °C. The prospect of a coolant at these temperatures circulating in the reactor vessel demands that careful analysis be performed to ensure that excessively hot spots are not created and that sufficient mixing of the coolant is obtained. Computational fluid dynamics (CFD) coupled with heat transfer will be used to perform the desired analyses. However, primarily because of the imperfect nature of modeling turbulent flow, any CFD calculations used to perform nuclear reactor safety analysis must be validated against experimental data. Experimental data have been taken in a scaled section of the lower plenum of a prismatic VHTR at the matched index of refraction (MIR) facility at the Idaho National Laboratory. These data were taken with the intent that they be examined for use as validation data. A series of investigations have been conducted to assess the MIR data. Issues that have already been examined include the extent of the required computational domain, the outlet boundary condition, the inlet data and the effect of the turbulence model. One of the jets that flow into the model impacts on a wedge, which represents a portion of a hexagonal graphite block that is part of the inner wall of the lower plenum. The nature of the flow below this particular jet is such that a randomly varying recirculation zone is created. This recirculation zone is seen to change in size, causing a relatively long-time scale of motion or disturbance on the flow downstream. It is concluded that such a feature is undesirable in a validation data set, firstly because of its apparent random nature and, secondly, because to obtain an appropriate longtime average would be impractical because of the compute time required. It is found that by eliminating the first of the four inlet jets into the scaled model, the resulting recirculation zone is rendered stable.

Topics: Flow (Dynamics)
Commentary by Dr. Valentin Fuster
2010;():517-523. doi:10.1115/ICONE18-29502.

In order to study the reverse flow characteristics in U-tubes of steam generator in the natural circulation case, the code RELAP5/MOD3.3 is used to model and calculate single-phase water flow for PWR under some typical operating conditions in the natural circulation case. The U-tubes of steam generator are classified according to their length and then are divided into different nodes and flow lines. The calculated results show that reverse flow exists in some inverted U-tubes of the steam generator, the natural circulation capacity of the primary coolant circuit system declines and the calculated net mass flux of the natural circulation accords with the experimental data. The traditional lumped parameter method can not simulate the reverse flow characteristics in inverted U-tubes and its result is much greater than the experimental data. When the steam generator outlet pressure is higher than inlet pressure, and gravitational pressure drop is lower than the total of frictional pressure drop and area change pressure drop, the reverse flow will occur. As to the nuclear power plant described in this paper, the mass flux of the shorter U-tubes drops more quickly and at last reverse flow will occur. The temperature distribution is uniform in inverted U-tubes, and it is almost identical with that of SG in secondary side. The occurrence of reverse flow can be judged by that whether the steam generator inlet temperature is lower than reactor outlet temperature obviously. It is indicated that reverse flow occurred in the U-tubes of the steam generator reduces the mass flux in the natural circulation system.

Commentary by Dr. Valentin Fuster
2010;():525-532. doi:10.1115/ICONE18-29516.

The concept of the liquid lithium target system for the boron neutron capture therapy (BNCT) using an accelerator was formulated. A lithium flow loop was designed and fabricated for development of lithium jet nozzles that provide a stable lithium plane jet. A water flow loop was designed, fabricated and used for development of lithium jet nozzles. Experiment of water plane jet was conducted to investigate the stability of high velocity plane jets. A large amount of droplets were entrained from the jet surface when the jet was unstable. The jet was observed by being illuminated with a stroboscope. It was found that a stable plane jet was realized by using a nozzle with the gap of 0.5mm and the length of 70mm. It was concluded that a fully developed flow without inlet disturbances at the outlet of the nozzle was required for the stability of the high velocity plane jet.

Commentary by Dr. Valentin Fuster
2010;():533-539. doi:10.1115/ICONE18-29522.

As a fundamental study for a direct contact type of lead-bismuth-cooled fast reactor (PBWFR), the method of gamma-ray radiography was applied to the visualization of the nitrogen bubbles in lead-bismuth eutectic (LBE) in a vertical square duct with inner size of 10×10×73.3 mm3 . Nitrogen gas was injected from a bottom nozzle with inner diameter of 3mm at the flow rate from 0.13 to 0.40NL/min. As γ–ray source, 60 Co with 11TBq was used, and a multi-crystal Gd2 O2 S (Tb) scintillator was employed for detection of γ–ray. The bubble behavior was visualized successfully by means of this method. Void fraction distributions were obtained by processing the visualized images, and the bubble length and bubbles rising velocity were determined from the result. There were some unstable bubble behavior depending on gas injection flow rate.

Commentary by Dr. Valentin Fuster
2010;():541-547. doi:10.1115/ICONE18-29523.

In order to clarify the fragmentation of molten core structural material (stainless steel) and molten metallic fuel and claddings on liquid phase formed by metallurgical reactions (liquefaction temperature = 650°C) during core disruptive accidents (CDAs), the present study focuses on the fragmentation of single molten stainless steel (316SS) and aluminum droplet penetrating a sodium pool. The temperatures of 3–5g molten aluminum droplets were 1002 to 1399°C, and the sodium pool was about 300°C. The instantaneous contact interface temperatures (Ti ) between the molten aluminum droplets and liquid sodium were calculated to be from 741°C below the boiling point of sodium (Tc,bp ) to 1019°C above Tc,bp . The temperatures of 5g molten 316SS droplets were 1510 to 1706°C, and the temperatures of sodium pool vary about 300–400°C. The Ti values between the molten 316SS droplets and liquid sodium were calculated to be from 916 to 1082°C. Fragmentation of the single molten aluminum droplet was clearly observed even at TiTc,bp . When Ti is approximately equal to or higher than the boiling point, the intensive fragmentation of droplet was clearly observed independent of Wea condition. Fragmentation of the single molten 316SS droplet was clearly observed even at Ti below its melting point. The Dm values of aluminum and 316SS droplets with relatively high Wea tend to be lower than those of droplets with relatively low Wea under the relatively low Ti condition. These results indicate the fragmentation of the molten core structural material and eutectic alloy fuels in liquid phase formed by the metallurgical reactions could possibly occur under the low Ti condition below and above the sodium boiling point, which is promising to assure the termination of accidents in CDAs and useful to the core design with enhanced safety in FBRs.

Commentary by Dr. Valentin Fuster
2010;():549-554. doi:10.1115/ICONE18-29531.

The CANDU 6 nuclear power plant is modeled with SCDAP/RELAP5 code including the heat transport system (HTS), pressure and inventory control system, calandria vessel (CV) and main steam lines, etc. Through simulation of station blackout (SBO), the results obtained before the fuel channel dry-out are illustrated. Before the steam generators (SGs) dry out, the core decay heat could be removed through the SG secondary side, and fuel temperature, pressure of the HTS could keep low. The loss of the HTS inventory results in the fuel channel dry-out, heat-up of the fuel, the calandria and pressure tubes. The moderator temperature remains low in the whole process.

Commentary by Dr. Valentin Fuster
2010;():555-560. doi:10.1115/ICONE18-29536.

An investigation is made on the thermal transfer characteristic of horizontal convection in the fuel transfer channel. The convection is a complex 3-D flow. Influencing factors on the convection are calculated and compared, such as the stuck position of fuel assembly, the diameter of channel, and the mechanical structure of the fuel transfer system. An optimized mechanical design for fuel transfer system is suggested. It seems effectively improved cooling effect by numerical simulation validation.

Commentary by Dr. Valentin Fuster
2010;():561-567. doi:10.1115/ICONE18-29541.

Fluctuating flow is widely presented in nuclear power plant operating procedure. When the fluctuating flow occurs in the loop, the fluid flow and heat transfer in the core will be affected, which makes the study of flow fluctuation have more practical significance. With computational fluid dynamics (CFD), characteristics of fluid flow and heat transfer are numerically simulated in a horizontal tube under periodical fluctuating flow. The influences of different factors on the fluid flow and heat transfer are analyzed. The simulation results of steady flow and heat transfer in horizontal tube agree with the traditional empirical correlations’ results, which validates the feasibility of doing this research using CFD simulation. The horizontal tube fluctuation flow and heat transfer with different flow fluctuation periods, fluctuation relative amplitudes and heat fluxes are numerically simulated. The results show that the smaller the flow fluctuation period is, the larger the flow fluctuation relative amplitude we get, and the more evident influence of flow fluctuation on fluid flow and heat transfer can be found. The larger the heat flux is, the larger amplitude of temperature fluctuation of fluid will be. What is more, there is a lag in phase between friction coefficient and velocity, which is not presented between heat transfer coefficient and velocity.

Commentary by Dr. Valentin Fuster
2010;():569-577. doi:10.1115/ICONE18-29569.

One of the strategies of cost reduction of nuclear power generation is the increase of power outputs. Especially, in order to achieve performance upgrade of Advanced Boiling Water Reactor (ABWR), it is extremely important to evaluate coolant flow in the lower plenum of ABWR. With the plenty construction in the lower plenum, it is thought that the flow structure is complicated. Moreover, according to the previous studies, there is strong evidence that vortexes arise around side entry orifice when coolant flows in there. Such complicated flow may affect the pressure loss (differential pressure in the lower plenum) and the coolant flow distribution to each core fuel assemblies, and consequently it would influence advancement of fuel economics. Although the simulation results by a CFD code can predict such complicated flow in the lower plenum, the accuracy of simulation data are not enough. Hence, the present study is focusing on the establishment of the benchmark of CFD code by using the visualization method in the lower plenum of ABWR. The objective of the present study is to investigate correlation between the structure of vortexes and complicated flow in upstream of core support beam, and the effect of such fluid behavior to the differential pressure. In the constructed model of the lower plenum of ABWR, velocity profiles were measured by LDV (Laser Doppler Velocimetry) and PIV (Particle Image Velocimetry) techniques. And differential pressure of constructed model is measured by differential pressure instrument. Each measurement was worked out in the range of Reynolds number from 103 to 104 . It was found from the LDV measurement that the velocity at the center of the test section was faster than that near the wall in upstream. In downstream, the velocity profiles showed the tendency to be flat in the core support beam. Vortexes were observed around side entry orifice by PIV measurement. Concerning differential pressure, it is necessary to examine correlation between complicated flow structure and differential pressure. Thus in the present study, the differential pressure distribution of constructed model is experimentally investigated.

Commentary by Dr. Valentin Fuster
2010;():579-584. doi:10.1115/ICONE18-29602.

In the Very-High-Temperature Reactor (VHTR) which is the next generation nuclear reactor system, ceramics and graphite are used as the fuel coating material and the core structural material, respectively. Even if the accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. This is because the thermal capacity of the core is so large. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). From the view point of the safety characteristic, the passive cooling system should be designed for the VHTR as the best way of the reactor and vessel cooling systems (VCS). So, the gas cooling system by natural convection is the one of the candidate systems for the VCS of the VHTR. This study is to develop the passive cooling system for the VHTR using the vertical rectangular channel inserting porous materials. In general, when the high temperature circular or rectangular channels are cooled by forced convection of gas, there are several methods for enhancement of heat transfer such as attaching radial or spiral fins on a channel surface or inserting twisted tape in a channel. The objective of this study is to investigate heat transfer characteristics by forced convection of porous materials inserted into a rectangular channel with high porosity. In order to obtain the heat transfer characteristics of the one-side heated vertical rectangular channel inserting the porous material, an experiment was carried out. From the results obtained in this experiment, it was found that an amount of removed heat by forced convection using porous material (porosity > 0.996) was about 15% higher than that without the copper wire. Furthermore, the ratio between the amounts of heat removed of the rectangular channel with the porous material and without the porous material increases with increasing temperature of the channel wall.

Commentary by Dr. Valentin Fuster
2010;():585-592. doi:10.1115/ICONE18-29617.

Sloshing dynamics of a molten core is one of the fundamental behaviors in core disruptive accidents of a liquid-metal cooled reactor. In addition, solid particle-liquid mixture comprising molten fuel, molten structure, refrozen fuel, solid fuel pellets, etc. could lead to damping of its flowing process in a disrupted core. The objective of the present study is to investigate the applicability of the finite volume particle method (FVP), which is one of the moving particle methods, to 3D motion of liquid sloshing processes measured in a series of experiments. In the first part of this study, a typical sloshing experiment of single liquid phase is simulated to verify the present 3D FVP method for sloshing characteristics that include free surface behaviors. Second, simulations of sloshing problems with solid particles are performed to validate the applicability of the FVP method to the 3D motion of solid particle-liquid mixture flows. Some good agreements between the simulation and its corresponding experiment demonstrate applicability of the present FVP method to 3D fluid dynamics of liquid sloshing flow with solid particles.

Commentary by Dr. Valentin Fuster
2010;():593-596. doi:10.1115/ICONE18-29619.

This article reviews the studies on steam generator tube rupture (SGTR) accident in lead alloy-cooled fast reactors (LFR) and the accelerator-driven transmutation system (ADS) including proposal for visualization test in injection of water into molten lead-bismuth eutectic (LBE) and lead (Pb) using neutron radiography. The SGTR and the consequent interaction between LBE/Pb and water, i. e., coolant-coolant interactions (CCI), represents an important concern for the safety of LFR and ADS since the CCI might bring a local core voiding, and even initiate a sloshing motion of liquid metal coolant to damage the vessel, which is similar with fuel-coolant interaction (FCI) in light water reactors (LWR) and sodium-cooled fast reactors. Based on the review of related studies, visualization test in special conditions using neutron radiography is proposed to investigate this transient thermal-hydraulic phenomenon.

Topics: Accidents , Boilers , Rupture
Commentary by Dr. Valentin Fuster
2010;():597-604. doi:10.1115/ICONE18-29630.

The flow-induced void fraction transition phenomenon was observed in an upward air-water two-phase flow in a vertical pipe with inner diameter D = 200 mm and height z = 25 m. As the two-phase flow develops in a vertical pipe, the void fraction increases firstly in the flow direction in bubbly flow, then decreases in the flow direction, finally increase again. The flow-induced void fraction transition shows an N-shaped changing manner. The present experimental investigation revealed that this phenomenon was attributed to the formation and the growth of local dominant large bubbles in the flow. According to the bubble sizes and behaviors observed in the experiment, the flow regimes were classified into bubbly, churn and slug flows in a vertical large-diameter pipe. The drift velocities in the three flow regimes were measured in this paper. New constitutive equation for drift velocities in bubbly, churn and slug flows was proposed and confirmed in this study. The flow-induced void fraction transition in N-shaped manner can be predicted by using the drift flux model with the newly developed constitutive equations.

Commentary by Dr. Valentin Fuster
2010;():605-613. doi:10.1115/ICONE18-29644.

Visualized experimental observation on flow patterns during flow boiling of water under single-side heated and fluid-inlet subcooled conditions in a vertical narrow rectangular channel with the cross-section of 40×3mm2 have been carried out. Four discernible flow patterns which names dispersed bubbly, coalesced bubbly, churn flow and annular flow are obtained. Flow visualization in two dimensions of two-phase flow patterns for narrow rectangular channel, which provided clearer evidence to distinguish flow patterns, have been performed. Based on the experimental results, a flow pattern map for single-side heated narrow rectangular channel has been developed and then compared with the exiting maps and flow transition criteria.

Commentary by Dr. Valentin Fuster
2010;():615-622. doi:10.1115/ICONE18-29659.

In this paper, numerical simulation of cavitation for water and sodium flows were performed by using CFD code package FLUENT and the results were compared with the experiments. The geometry of the grid for numerical calculation was a venturi with ID and OD of 6.5 and 21 mm, respectively. The numerical simulations showed that the onset cavitation conditions were affected significantly by the non-condensable gas content in the liquid. Comparing with the experimental results, the non-condensable gas contents that gave close predictions of onset cavitation conditions with the experiments were larger in water than in liquid sodium because of the different solubility property of air and argon gas. Parametrical analysis showed that the void fraction distributions for water and liquid sodium did not show large differences in case of the same caviation coefficient and non-condensable gas contents.

Commentary by Dr. Valentin Fuster
2010;():623-628. doi:10.1115/ICONE18-29671.

Based on a heuristic hypothesis that alkaline metals with one single electron on the outermost shell would little interact with externally applied RF field above plasma frequency, a rudimentary experiment as well as the theoretical estimate of the energy structure been performed to further explore the spectroscopic properties of liquid sodium (Na)[1–3] . Consequently, it was successfully proven that Na is reasonably transparent to the VUV (vacuum ultraviolet) laser radiation, although the liquid Na surface is highly reflective, being like a mirror to human eyes. The impact of this result is that the velocity field information inside the liquid Na can be visualized by implementing the well developed PIV (particle image velocimetry) technique[4, 5] . A large eddy simulation (LES) code has also been developed for comparison with the experimental results[6] . Furthermore, the newly developed Na loop is designed so as to enable the application of electric and magnetic field in the orthogonal direction to each other that vigorous dynamics of vortices inside the liquid Na are resolved in the phase space under the Lorentz force. The results herein obtained contribute not only to the thermal hydraulics in fast reactors[7, 8] but also space physics, such as the spiral galaxy formation[9] and solar flare activities[10] .

Commentary by Dr. Valentin Fuster
2010;():629-634. doi:10.1115/ICONE18-29679.

A two-dimensional code was developed to simulate vortex shedding characteristic and flow-structure interaction (FSI) of plate-type structures. In the code the physical component boundary fitted coordinate (PCBFC) was used to deal with the curve boundary. The arbitrary Lagrangian Eulerian (ALE) method was used to realize the grid movement. A barrier unit idea was adopted to deal with the boundary of fluid domain and solid domain in the code. The code was validated by comparing the numerical simulation results with experimental data. It was found that the vortex shedding phenomena in case of rectangular cylinder are strongly related to the length of the rectangular cylinder in the stream line.

Commentary by Dr. Valentin Fuster
2010;():635-644. doi:10.1115/ICONE18-29687.

The purpose of this paper is to develop a nonlinear model to investigate the instabilities of a two-phase natural circulation loop under low-pressure condition. Inlet velocity oscillations and the corresponding trajectories are respectively presented in the time evolution planes and phase planes. We obtain a stability map to explore the instability regions of this natural circulation loop. The results show that the considered loop has two unstable regions, instability type-I in the low power region and instability type-II in the high power region. Then the parametric study is carried out to understand the relation between the parameters of system and two types of instability. The parametric study reveals that lengthening the riser has an unstable effect on system stability. Thus, lengthening the riser causes a reduction in the stability region in the both low power and high power levels. Also it can be observed that by increasing the form loss coefficient at the inlet of heated section or in the downcomer section, the stability region expands, however by increasing the form loss coefficient at the outlet of heated section or in the upper horizontal section, the stability region decreases consequently.

Topics: Pressure
Commentary by Dr. Valentin Fuster
2010;():645-654. doi:10.1115/ICONE18-29688.

Two broken control rods and a large number of rods with cracks were found at the inspection carried out during the refueling outage of the twin reactors Oskarshamn 3 and Forsmark 3 in the fall of 2008. As a part of an extensive damage investigation, time dependent CFD simulations of the flow and the heat transfer in the annular region formed by the guide tube and control rod stem were carried out, [1]. The simulations together with metallurgical and structural analyses indicated that the cracks were initiated by thermal fatigue. The knowledge assembled at this stage was sufficient to permit the restart of both reactors at the end of year 2008 conditioned to that further studies to be carried out for clarifying all remaining matters. Additionally, all control rods were inserted 14% to protect the welding region of the stem. Unfortunately, this measure led to new cracks a few months later. This matter will be explained in the second part of this work, [2]. As a part of the accomplished complementary work, new CFD models were developed in conformity with the guidelines of references [3] and [4]. The new results establish the simulation requirements needed to accomplish accurate conjugate heat transfer predictions. Those requirements are much more rigorous than the ones needed for flow simulations without heat transfer. In the present case, URANS simulations, which are less resource consuming than LES simulations, seem to rather accurately describe the mixing process occurring inside the control rod guide tube. Structure mechanics analyses based on the CFD simulations show that the cracks are initiated by thermal fatigue and that their propagation and growth are probably enhanced by mechanical vibrations.

Commentary by Dr. Valentin Fuster
2010;():655-663. doi:10.1115/ICONE18-29689.

A large number of control rod cracks were detected during the refuelling outage of the twin reactors Oskarshamn 3 and Forsmark 3 in the fall of 2008. The extensive damage investigation finally lead to the restart of both reactors at the end of 2008 under the condition that further studies would be conducted in order to clarify all remaining matters. Also, all control rods were inserted 14% in order to locate the welding region of the control rod stem away from the thermal mixing region of the flow. Unfortunately, this measure led to new cracks a few months later due to a combination of surface finish of the new stems and the changed flow conditions after the partial insertion of the control rods. The experimental evidence reported here shows an increase in the extension of the mixing region and in the intensity of the thermal fluctuations. As a part of the complementary work associated with the restart of the reactors, and to verify the CFD simulations, experimental work of the flow in the annular region formed by the guide tube and control rod stem was carried out. Two full-scale setups were developed, one in a Plexiglass model at atmospheric conditions (in order to be able to visualize the mixing process) and one in a steel model to allow for a higher temperature difference and heating of the control rod guide tube. The experimental results corroborate the general information obtained through CFD simulations, namely that the mixing region between the cold crud-removal flow and warm by-pass flow is perturbed by flow structures coming from above. The process is characterized by low frequent, high amplitude temperature fluctuations. The process is basically hydrodynamic, caused by the downward transport of flow structures originated at the upper bypass inlets. The damping thermal effects through buoyancy is of secondary importance, as also the scaling analysis shows, however a slight damping of the temperature fluctuations can be seen due to natural convection due to a pre-heating of the cold crud-removal flow. The comparison between numerical and experimental results shows a rather good agreement, indicating that experiments with plant conditions are not necessary since, through the existing scaling laws and CFD-calculations, the obtained results may be extrapolated to plant conditions. The problem of conjugate heat transfer has not yet been addressed experimentally since complex and difficult measurements of the heat transfer have to be carried out. This type of measurements constitutes one of the main challenges to be dealt with in the future work.

Commentary by Dr. Valentin Fuster
2010;():665-672. doi:10.1115/ICONE18-29690.

Ultrasonic technique was widely applied for the measurement of two-phase flow, however, the majority techniques deal primary with pulse echo technique under higher operating sound frequency in a MHz order since higher accuracy can be obtain for the dynamic gas-liquid interface locations. On the other hand the transmission type ultrasonic methods was also developed for bubbly flow diagnostics for the time and cross sectional averaged void fraction measurement or bubble two-phase flow parameter determination by pulse echo techniques with higher operating sound frequency in a MHz order. In this work, the transmission type ultrasonic imaging of horizontal two-phase flow was experimentally investigated. Ultrasonic transducer used is 150 kHz and time averaged void fraction results for stratified flow was compared with capacitance void measurement and ultrasonic pulsed echo techniques. The results show that the transmission intensity for the stratified flow pattern decreases with increasing void fraction until 55% of void then increases with increasing void fraction till 100%. An application to plug, slug and annular flow regime also investigated in detail.

Commentary by Dr. Valentin Fuster
2010;():673-679. doi:10.1115/ICONE18-29695.

This paper reports an experimental study of the measurement of elongated bubbles velocities and their longitudinal shapes using a high speed ultrasonic system in concurrent horizontal and at 5° and 10° inclined upward flow. The circular pipe test section is made of 25.6 mm stainless steel, followed by a transparent acrylic pipe with the same diameter. The high speed ultrasonic system consists of two transducers (10 MHz/6.35 mm diameter), a generator/multiplexer board that convert analog signals into digital data at a rate of 100 million frames per second, and a software that stores all the frames and the results of the time of flight of each signal. The results are compared with a visualization technique that consists of a high-speed digital camera recording images at rates of 125 and 250 frames per second. This range of liquid superficial velocity is from 0.2 to 1.1 m/s and that of the gas superficial velocity is from 0.35 to 1.0 m/s. The results obtained with the two experimental techniques show a good agreement among them for the elongated bubbles lengths and velocities, while having great statistics dispersion. The measured bubble shape is in agreement with literature data.

Commentary by Dr. Valentin Fuster
2010;():681-688. doi:10.1115/ICONE18-29704.

Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can achieve a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) The pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump. (3) The wire wrap pitch has significant effect on the flow in the assembly. Smaller Hwire /Drod will result in stronger cross flow and a more uniform coolant temperature profile, and also a higher pressure drop.

Commentary by Dr. Valentin Fuster
2010;():689-696. doi:10.1115/ICONE18-29707.

An artificial neural network (ANN) for predicting critical heat flux (CHF) of concentric-tube open thermosyphon has been trained successfully based on the experimental data from the literature. The dimensionless input parameters of the ANN are density ratio, ρl /ρv , the ratio of the heated tube length to the inner diameter of the outer tube L/Di , the ratio of frictional area, di /(Di + do ), and the ratio of equivalent heated diameter to characteristic bubble size, Dhe /[σ/g(ρl −ρv )]0.5 , the output is Kutateladze number, Ku. The predicted values of ANN are found to be in reasonable agreement with the actual values from the experiments with a mean relative error (MRE) of 8.46%. For a particular outer tube, the CHF increases initially and then decreases with increasing inner tube diameter, and has a maximum at an optimum diameter of inner tube (do,opt ). The do,opt is correlated with the working fluid and may decrease with the increase of ρl v . CHF decreases with the increase of L/Di , and the decreasing rate decreases as L/Di increases. In the influence scope of pressure, the CHF decreases with increasing pressure for R22, while increases with increasing pressure for R113.

Commentary by Dr. Valentin Fuster
2010;():697-704. doi:10.1115/ICONE18-29710.

In a core disruptive accident of a fast breeder reactor, the post accident heat removal is crucial to achieve in-vessel core retention. Therefore, a series of experiments on bubble behavior in a particle bed was performed to clarify three-phase flow dynamics in debris bed, which is essential in heat-removal capability, under coolant boiling conditions. Although in the past several experiments have been carried out in the gas-liquid-solid system to investigate the bubble dynamics, most of them were under lower solid holdup (≤ 0.5), where the solid-phase influence may be not so important as much as the liquid phase. While for this study, the solid holdup is much higher (> 0.55) where the particle-bubble interaction may be dominated. The current experiment was conducted in a 2D tank with the dimensions of 300 mm height, 200 mm width and 10 mm gap thickness. Water was used as liquid phase, while bubbles were generated by injecting nitrogen gas from the bottom of the tank. Various experimental parameters were taken, including different particle bed height (from 30 mm to 200 mm), various particle diameter (from 0.4 mm to 6 mm), different particle type (acrylic, glass, alumina and zirconia beads), and different nitrogen gas flow rate (around 1.75 ml/min and 2.7 ml/min). By using digital image analysis method, three kinds of bubble rise behavior were observed under current experimental conditions and confirmed by the quantitative detailed analysis of bubble rise properties including bubble departure frequency and bubble departure size. This experiment is expected in the future to provide appropriate quantitative data for the analysis and verification of SIMMER-III, an advanced fast reactor safety analysis code.

Commentary by Dr. Valentin Fuster
2010;():705-713. doi:10.1115/ICONE18-29719.

Regarding the Japan Sodium-cooled Fast Reactor, a multi-elbow piping system is adopted for its cold-legs. Flow Induced Vibration (FIV) is considered to be caused by complex flow with very high velocity in the elbows. In this study, pressure measurement test of a single elbow flow is conducted to find out pressure fluctuation characteristic which is related to the elbow turbulent flow and lead potentially to the FIV. Two types of experimental loops, that is, 1/7 and 1/15-scale setup simulating the JSFR cold-leg pipings, are used for pressure measurement, and a distinguishing peak can be seen in the power spectrum density profile of pressure fluctuation obtained where flow separation occurs and at the downstream from it. This characteristics of pressure fluctuation is obtained from the two different scale experiments, and the scale effect is not found in terms of the pressure fluctuation.

Commentary by Dr. Valentin Fuster
2010;():715-722. doi:10.1115/ICONE18-29725.

The two Loviisa VVER-440 type reactors were commissioned in 1977 and 1980. The original designed life time of the reactors was 30 years. In 2003 Fortum, the owner and the operator of the Loviisa plant, launched an extensive safety study to prove the authorities that there was not any major safety issue why operating license could not be extended for another 20 years. In 2007 the Ministry of Employment and the Economy of Finland granted 20 and 23 years extension to the operating license for units 1 and 2, respectively. One issue, which needed further investigation, was the core cooling capability during sump circulation; i.e. were the present sump strainers good enough to prevent insulation fiber from not clogging the core coolant flow? Back in the 1990’s the original steel wire type sump strainers were replaced with stronger steel pipe type strainers. Some time later experiments were carried out to find out if insulation fiber could penetrate through the strainer holes and reduce the coolant mass flow rate through the core. The experiments indicated that the insulation fiber mixed with coolant partly penetrates through the strainer and gathers to the fuel assembly spacer grids increasing pressure loss across the core. The experiments were carried out in a rather simple test facility and also under forced single phase circulation. In those loss-of-coolant accidents (LOCA) where sump circulation takes place, circumstances are completely different. Therefore, it was decided that the APROS (Advanced PROcess Simulation) simulation software would be used to study the insulation fiber effect on core coolability during the accident. A large LOCA was chosen for the case to be analyzed. The reason for this was that during a large LOCA sump circulation begins in the early phase of the accident and a lot of emergency core cooling (ECC) water is injected into the primary circuit during sump circulation. The paper will first shortly discuss APROS simulation software. Then the test facility and the experimental results will be presented. The main issue is the analyses results. Several analyses were carried out to be able to determine the maximum amount fiber gathered in the spacer grids which the core can tolerate without overheating.

Commentary by Dr. Valentin Fuster
2010;():723-727. doi:10.1115/ICONE18-29731.

The applicability of artificial roughness in light-water reactors is investigated for the purpose of heat transfer improvement in fuel rod bundles. Since the roughening technique has a significant impact on friction losses, the investigation is divided in two distinct steps: flow resistance and convective heat transfer. The present paper deals with roughness effects on flow resistance. The technique consists of a multiplicity of small elements distributed on the surface of the simulated fuel rod. A parallel rib-type roughness is selected for this work for simplicity and since it has been extensively investigated in the past. Locally flow resistance is simulated using Computational Fluid Dynamics, CFD, in smooth and in rough rod bundles downstream of support grids with and without flow-enhancing features (vanes). This investigation is performed with basis on experimental testing. With model parameters established, various candidate roughness designs can evaluated for minimum flow resistance.

Commentary by Dr. Valentin Fuster
2010;():729-736. doi:10.1115/ICONE18-29765.

One of the limiting conditions during operation of a Pressurized Water Reactor is cladding integrity in case of occurrence of any conditions I or II events. The decoupling criterion is the absence of Departure from Nucleate Boiling (DNB) during the full sequence of any of these transients. Heat transfer between the clad and the water is limited by the DNB phenomenon when local surface heat flux is greater than the so-called Critical Heat Flux (CHF). Heat production at the surface is higher than heat removal capacity by the coolant therefore a vapor blanket is formed around the clad; consequently the heat transfer will drastically drop resulting in a sudden significant increase of the local wall temperature and clad damage may appear if no corrective action is initiated. DNB can not be estimated with physical principles only. Experimental support is needed for evaluation. Occurrence of DNB is evaluated using the Departure from Nucleate Boiling Ratio (DNBR) which is a function of both core thermal hydraulic (T/H) parameters and design of the fuel assembly. Advanced fuel assemblies claim higher CHF values compared to previous designs. Along with increased DNB performances for advanced fuel assemblies, CHF correlation development and advanced methodologies enable to extend normal operating conditions of a nuclear plant. On the one hand, CHF performances really increased allow additional margin related to the loss of fuel cladding integrity whereas on the other hand optimized correlations and advanced methodologies reduce this margin. An accurate assessment of the CHF performance of the advanced fuel assemblies is therefore required. This paper will raise issues regarding the assessment of the CHF performance of new advanced fuel assemblies design. The issues will be focused on the reliability of the experimental assessment of the CHF values and the accuracy of the transposition of mock up geometries to plant core configuration (representativity of the experiments). The verification that the tests conditions (pressure, flowrate, quality, heat flux [[ellipsis]]) ensure a proper coverage of all core conditions encountered during any of the conditions I & II transients is closely linked to DNBR methods and will not be extensively covered in this paper. This paper suggests some thoughts about relevance of the demonstration carried out by vendors on these matters.

Commentary by Dr. Valentin Fuster
2010;():737-742. doi:10.1115/ICONE18-29771.

A Lungmen RETRAN-3D model has been constructed to predict the transient behaviors for startup test, furthermore verify the acceptance criteria specified in the documents of startup test procedure. This study focuses on the prediction of the startup test with Load Rejection (LR) with bypass and the parametric analysis of lead-lag time constants in pressure regulator. For the analysis of LR with bypass, the major mitigation functions, i.e., Selected Control Rods Run-In (SCRRI) and turbine bypass function, are simulated to examine whether scram is initiated during the transient or not. The analytic results show the reactor is brought to a steady state without scram. The neutron flux in the final state is around 34%, and the pressure regulator sensed maximum pressure rise is limited to a maximum of 3kPa. The result also shows that the 110% steam bypass capacity is capable to mitigate the power increase caused by the positive reactivity insertion as a result of pressure-wave-induced void collapse. For the parametric analysis of lead-lag time constants in pressure regulator, the time domain response of Steam Bypass and Pressure Control System (SBPCS) is demonstrated by a step change of pressure setpoint and different combinations of lead-lag time constants defined in pressure regulator. The results show that the responses, i.e., response time and overshooting, are minimized when the lag time constant is between 4 to 6 seconds and the lead time constant is 50% to 70% of lag time constant. The analysis result of SBPCS provides the trend as a reference for the adjustment of lead-lag time constants during the future Lungmen startup test.

Commentary by Dr. Valentin Fuster
2010;():743-750. doi:10.1115/ICONE18-29794.

It is very important to study bubble growth and departure from the nucleation site for better understanding of boiling heat transfer in a narrow channel. Bubble growth and departure in a narrow rectangular under atmosphere pressure is visually observed by the wide and narrow side of the narrow rectangular channel using high speed digital camera. There is a small bubble contact diameter between the bubble base and heating surface when the bubble is growing at the nucleation site, and the growing bubble shape is almost spherical. The bubble growth law at the different nucleation sites is almost uniform under the condition of the same thermal parameters, but bubble departure diameters are obvious distinct because of different sizes of nucleation sites. In the current study, the bubble growth rate in a narrow rectangular channel is small, and the bubble departure time is long, the bubble growth diameter can be predicted by using the amendatory Zuber expression. The effect of thermal parameters on the mean bubble departure diameters is statistical analysed in the view window, the mean bubble departure diameters decrease with increasing heat flux, the mean bubble departure diameters decrease with increasing inlet subcooling, the mean bubble departure diameters decrease with increasing bulk flow velocity.

Commentary by Dr. Valentin Fuster
2010;():751-759. doi:10.1115/ICONE18-29805.

Liquid sloshing phenomenon is encountered whenever a liquid in a container has an unrestrained surface and can be excited. Specific type of sloshing motion can occur during the core meltdown of a liquid metal reactor (LMR) and can lead to a compaction of the fuel in the center of the core and to energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomenon a numerical method should be capable to predict motion of free surface of liquid, wave propagation and reflection from the walls. A meshless method based on Smoothed Particle Hydrodynamics (SPH) for the simulation of 3D free surface liquid motion has been developed. Proposed method is applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results as well as with results of computations performed with 3D code SIMMER-IV which is an advanced reactor safety analysis code that implements the traditional mesh-based numerical method. In series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.

Commentary by Dr. Valentin Fuster
2010;():761-768. doi:10.1115/ICONE18-29817.

A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.

Commentary by Dr. Valentin Fuster
2010;():769-775. doi:10.1115/ICONE18-29823.

Spiral-fin fuel elements belong to the self-spacer fuel elements; they are usually arrayed in hexagonal fuel assembly and widely used in the core of the tight lattice reactor. Closely packed and helical spacer makes the heat transfer course and phenomena in the channel of the spiral-fin fuel bundle very complicated, with little literature being found in this field. The paper preliminarily studied the heat conduction through the finned can together with convection heat transfer between the fuel element and coolant. Modified fin efficiency was introduced to calculate the heat conduction; for the convention heat transfer, an imaginary channel named equivalent annulus was adopted. The geometric and physical influence of the spiral fins to the coolant in the annulus was studied. Comparing with annulus have inner surface being heated, it was found that the heated perimeter and velocity of coolant in the equivalent annulus were changed by the helical spacer, a new model correlation based on the equivalent annulus’ method was obtained by modifying the heat transfer correlation in annulus. Quantitative study was conducted for a given rods bundle, the results shows that the spiral-fins’ effect to the heat transfer is related to the change of the convection coefficient and heat conduction amount caused by the spacer, at the common condition, its impact is little.

Topics: Heat transfer , Fuels
Commentary by Dr. Valentin Fuster
2010;():777-783. doi:10.1115/ICONE18-29836.

Each BWR fuel design requires a method to predict its dryout performance in order to be licensed. Presently, the assessment of dry-out risk is based on empirical correlations, which sometimes results in inaccurate or non-physical predictions in certain portions of operational space. This poses a number of limitations as plant operators seek to extract additional value from the fuel through more aggressive operation strategies. A new form of BWR dryout correlation is developed. Accuracy of predictions outside of experimental data range is increased by employing a non-linear correlation form and the transformation to axial power profile, which is based on physical considerations. Proper qualitative behavior is assured by the correlation form itself rather than values of regression coefficients.

Commentary by Dr. Valentin Fuster
2010;():785-791. doi:10.1115/ICONE18-29840.

The projected prismatic and pebble-bed preconceptual designs for the Very High Temperature Reactor (VHTR) both have large diameter, horizontally-oriented, exit ducts that lead hot helium from the core exit through the lower plenum to the balance-of-plant. In the event a significant leak is present in the exit duct the system will depressurize and the possibility of air ingress into first the lower plenum and then the core must be considered. The intrusion of air against helium flow is driven by the weight difference between the two fluids. Thus the potential for air intrusion always exists. Whether such a phenomenon will actually occur is not as clear. This paper proposes a simple criterion for the onset of air intrusion. The criterion is expressed in terms of a densimetric Froude number (the inverse of the bulk Richardson number) and a scaled pressure difference between the duct exit and the containment. The pressure difference is rendered dimensionless by dividing it with the product of the duct radius and the difference of specific weight of the two fluids. The dimensionless variables in the criterion enable its validity be verified by experiments using fluids other than helium and air. This paper presents the development of this criterion and experimental validation using water and biodiesel as the testing fluids. The relevance and application of this criterion to VHTR air ingress scenarios are described.

Commentary by Dr. Valentin Fuster
2010;():793-799. doi:10.1115/ICONE18-29842.

The transient of Large Load Demand Change imposed on the Master Controller of the Recirculation Flow Control System is studied in this paper. The simulation performed in this study used the Core Design and Safety Analysis Package, which includes a lattice calculation code, CASMO-4, a three-dimensional core simulation code, SIMULATE-3, an interface code, SLICK, and a thermal hydraulic code, RETRAN. In the analysis the initial condition of 100% rated power/85% rated core flow was selected as base case, and the load demand was first manually ramped down from 100% to 75% (25% change) at the rate of 1%/sec and then ramped up at the same rate to the original level 150 seconds later. This paper presents the wind-up phenomenon observed in the output of the Proportional Integral (P/I) element when a large load demand change is imposed on the Master Controller of the RFCS. This wind-up phenomenon will cause an unnecessary time delay in response to the change to restore the Master Controller output to above the effective level. The sensitivity studies were conducted to get more insight into the wind-up phenomenon, which included: (1) load demand change of 20% level, (2) different initial core flow, and (3) different gain setting of P/I Controller. The analysis results show that the wind-up phenomenon can be avoided for the load demand change less than 20% or the higher initial core flow than 85% rated. Furthermore, the base setting of the P/I element of the Master Controller is appropriate to prevent the neutron flux from exceeding the scram setpoint with at least 10% margin during the ramp-up stage of load demand change.

Commentary by Dr. Valentin Fuster
2010;():801-807. doi:10.1115/ICONE18-29859.

The flowing characteristics of turbulent flow in rectangular channels in rolling and heaving motions are investigated theoretically with FLUENT code. The flowing model of turbulent flow in rectangular channels in rolling and heaving motion is established. The effects of several turbulent models and parameters on the flow are analyzed. In rolling motion, the velocity profile in channel center is more averaged. The mixing coefficient can be strengthened significantly in rolling motion. The effect of heaving motion on turbulent flow is more limited than that of rolling motion. The turbulent kinetic energy and fluctuating velocity in rolling motion are greater than that in heaving motion and steady state. The effect of rolling motion on the turbulent flow can be depressed by the tube wall.

Commentary by Dr. Valentin Fuster
2010;():809-814. doi:10.1115/ICONE18-29861.

The flowing and heat transfer characteristics of turbulent flow in tubes in rolling motion are investigated theoretically. The flowing and heat transfer models of turbulent flow in rolling motion are established. The correlations of frictional resistance coefficient and Nusselt number are derived. The results are also validated with experiments. The effects of several parameters on Nusselt number are investigated. The oscillating amplitude of Nusselt number is in direct ratio with Prandtl number and rolling frequency approximately. The more the flowing velocity is, the less the effect of rolling motion on the flow is. The variation of initial phase difference between Nusselt number and rolling motion with rolling frequency is very limited.

Commentary by Dr. Valentin Fuster
2010;():815-821. doi:10.1115/ICONE18-29867.

External reactor vessel cooling (ERVC) of the In-vessel retention (IVR) system is widely accepted as a feasible way to remove decay heat from the lower head of the reactor pressure vessel (RPV) under severe accident (SA) conditions. However, some issues relating to ERVC still need to be evaluated before its application, such as boiling and flow phenomena and CHF prediction, etc. To study these key issues, an experimental study program named REPEC (Re actor P ressure Vessel E xternal C ooling) is performed at Shanghai Jiao Tong University. Steady state experiments focusing on flow boiling phenomena investigation are carried out with comprehensive measurements, including temperature distribution, pressure drop and mass flow rate. As a part of studies on boiling mechanism and flow phenomena between RPV and the insulation, the experiment is analyzed and simulated with RELAP code. The code simulation covers most of the experimental cases, and a comparison between simulation results and experimental data are presented and discussed.

Commentary by Dr. Valentin Fuster
2010;():823-828. doi:10.1115/ICONE18-29870.

This paper briefly introduces recent progress in thermal-hydraulic R&Ds, which is mainly being performed at KAERI, for the APR+ (Advanced Power Reactor plus) development. The main R&D items for the APR+ reactor are associated directly with recent efforts to introduce new safety concepts in the APR+ standard design developments, which is currently in progress in the Republic of Korea. The R&D activities reported here mainly cover the thermal-hydraulic and severe accident areas and are being performed in experimental and/or analytical ways. They include: (1) advancement and optimization of safety injection system, (2) incorporation of passive safety features, such as advanced Fluidic Device (FD+) and passive auxiliary feedwater system (PAFS), and (3) incorporation of severe accident mitigation features.

Commentary by Dr. Valentin Fuster
2010;():829-834. doi:10.1115/ICONE18-29874.

PHENIX, a prototype sodium-cooled fast reactor (SFR), has demonstrated a fast breeder reactor technology and also achieved its important role as an irradiation facility for innovative fuels and materials. In 2009 PHENIX reached its final shutdown and the CEA launched a PHENIX end-of-life (EOL) test program, which provided a unique opportunity to validate an SFR system analysis code. The Korea Atomic Energy Research Institute (KAERI) joined this program to evaluate the capability and limitation of the MARS-LMR code, which will be used as a basic tool for the design and analysis of future SFRs in Korea. For this purpose, pre-test analyses of PHENIX EOL natural circulation tests have been performed and one-dimensional thermal-hydraulic behaviors for these tests have been analyzed. The natural circulation test was initiated by the decrease of heat removal through steam generators (SGs). This resulted in the increase of intermediate heat exchanger (IHX) secondary inlet temperature, followed by a manual reactor scram and the decrease of secondary pump speed. After that, the primary flow rate was also controlled by the manual trip of three primary pumps. For the pre-test analysis the Phenix primary system and IHXs were nodalized into several volumes. Total 981 subassemblies in the core were modeled and they were divided into 7 flow channels. The active 4 IHXs were modeled independently to investigate the change of flow into each IHX. The cold pool was modeled by two axial nodes having 5 and 6 sub-volumes, respectively. The reactor vessel cooling system was modeled to match the flow balance in the primary system. The flow path of vessel cooling system was quite complicated. However, it is simplified in the modeling. For a MARS-LMR simulation, the dryout of SGs have been described by the use of the boundary conditions for IHTS as a form of time-to-temperature table. This boundary condition reflects the increase in IHTS temperature by SG dryout during the initial stage of the transient and the increase in heat removal by the opening of the two SG containments at 3 hours after the initiation of the transient. Through the comparison of the pre-analysis results with the prediction by other computer codes, it is found that the MARS-LMR code predicts natural circulation phenomena in a sodium system in a reasonable manner. The final analysis for validation of the code against the test data will be followed with an improved modeling in near future.

Commentary by Dr. Valentin Fuster
2010;():835-847. doi:10.1115/ICONE18-29884.

One of the main concerns for modular Very High Temperature Gas-Cooled Reactors (VHTR) is the design of passive heat removal systems from the reactor vessel cavity. The Reactor Cavity Cooling System (RCCS) is an important heat removal system during normal and up-normal conditions. The design and validation of the RCCS is necessary to demonstrate that HTGRs can survive the postulated accidents. Here we investigate this using the Computational Fluid Dynamics (CFD) STAR-CCM+ V3.06.006 code to simulate the Pressurized Conduction Cooling (PCC) and Depressurized Conduction Cooling (DCC) accident scenarios. Heat is transported by radiation and free convection from the Reactor Pressure Vessel surface to the cooling panels or standpipes. The standpipes are cooled by natural circulation of air or forced circulation of water flowing through the pipes. A representative VHTR RCCS configuration was considered, represented experimentally by a 180° scaled model facility that was used to measure temperature and velocity distributions inside the cavity. The CFD model constructed incorporated the features of the experimental facility. Using the vessel temperature profile obtained from the experimental facility as boundary conditions in the CFD simulations, different tests were performed increasing the vessel average wall temperature progressively. Grid independence was achieved and different turbulence models and near-wall treatments were tested. For the standpipes, simulations with both natural circulation of air and forced circulation of water were performed. A reasonable agreement between the experimental results and the CFD simulations was achieved for the temperature distributions in the RCCS cavity. Also the standpipes external wall temperature was close to the experimental data. The fraction of heat exchange due to radiation determined by STAR-CCM+ code was in reasonable agreement with the experimental results. The k-ε turbulence models results were compared against the other turbulence models (i.e., the k-ω, Reynolds Stress Transport, and Spalart-Allmaras). Some differences were found between the turbulence models used. The k-ε turbulence models showed in general better performance than the k-ω and Spalart-Allmaras models if compared with the Reynolds Stress Transport (RST) results and experimental data. Among the k-ε turbulence models, the Realizable k-ε turbulence models with two-layer all-y+ near wall treatment performed better than the standard and the Abe-Kondoh-Nagano (AKN) k-ε models with Low-Reynolds Number low-y+ and all-y+ near wall treatments, if compared to both the RST and experimental results. The RST model was expected to perform better than the other models considering the strong anisotropy of the Reynolds stress tensor close to the vessel wall. The discrepancy between the experimental data with the RST model predictions may be due to the need for finer computational mesh model. A scaling analysis was developed to address the distortion introduced by the experimental facility and CFD model in simulating the physics inside the RCCS system with respect to the real plant configuration. The scaling analysis demonstrated that both the experimental facility and CFD model give a satisfactory reproduction of the main flow characteristics inside the RCCS cavity region, with convection and radiation heat exchange phenomena being properly scaled from the real plant to the model analyzed.

Commentary by Dr. Valentin Fuster
2010;():849-853. doi:10.1115/ICONE18-29885.

It is an important achievement of modern techniques to determine the mass flow rate and the phase fraction of wet steam by measuring the orifice plate differential pressure noise. The orifice plate differential pressure noise of air-water two-phase flow in horizontal and vertical rising pipelines were analyzed. Kinds of calculation methods were tried to get the differential pressure noise. From the difference waveform of the differential pressure square root that the acquisition card got and the mean square root of the sample that got before, the first in first out (FIFO) principle was used to get the differential pressure noise. Result shows that the differential pressure noise has different level at different vapor flow rate with the same water flow rate, conclusions show that the two-parameter measurement by using orifice plate differential pressure noise may be possibly used in vertical rising gas-water two phase flow.

Commentary by Dr. Valentin Fuster
2010;():855-863. doi:10.1115/ICONE18-29886.

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include 1) fuel pin failure and disruption, 2) molten pool boiling, 3) melt freezing and blockage formation, 4) duct wall failure, 5) low-energy disruptive core motion, 6) debris-bed coolability, and 7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt-solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and failure criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the specific phenomena of CDAs in SFRs in details.

Commentary by Dr. Valentin Fuster
2010;():865-872. doi:10.1115/ICONE18-29887.

Evaluation of wastage rate for nuclear power plant piping is important in maintaining plant reliability and safety. The two commonly well known causes of pipe wastage are liquid droplet impingement erosion (LDIE) and flow accelerated corrosion (FAC). This study is to evaluate the wastage rate of pipe wall and to measure the loss of mass caused by LDIE. A high speed rotating disk and water jet is used to perform LDIE phenomena. In this study, the liquid droplet impinging tests were conducted with the test piece mounted on the high speed rotating disk which crosses thin water jet. The amount of the wastage by LDIE was evaluated by changing the rotational speed, test time and test piece materials. After the experiment, the result of erosion was investigated by observing the surface of the test piece using a digital microscope, and photo was taken. A method of converting the photo into data is used. In this study, LDIE is considered as the main cause of test piece mass difference before and after the experiment. Therefore, a method is designed to evaluate the mass difference. In this method, the difference in volume of the material is calculated. The loss of mass is obtained by multiplying the difference in volume and density of the material. Then, 2-dimensional and 3-dimensional graphs of the test piece are drawn. The analytical code TRAC is used to analysis the liquid droplet velocity in Advance Boiling Water Reactor (ABWR) drain pipe.

Topics: Erosion , Pipes , Mechanisms
Commentary by Dr. Valentin Fuster
2010;():873-880. doi:10.1115/ICONE18-29895.

Among the existing CHF models, the bubble crowding model and the liquid sublayer dryout model have been well accepted for subcooled flow boiling. But both of the two models couldn’t give explanation about some details in the boiling crisis phenomenon according to photographic result. The aim of the present paper is to provide an improved synthesized model containing the characteristic of the above two models and then to give a comprehensive explanation about CHF. In the present model, the conservation equations of mass and energy are solved to derive the CHF formula. The length and velocity of the vapor blanket and the thickness of the liquid sublayer are needed. The quality and void fraction in bubble region and the core region are calculated by a homogeneous assumption. The vapor blanket length is thought to be equal to the Helmholtz wavelength and it is obtained from several parameters in the bubble region. The velocity of the vapor blanket is connected to the flow velocity of the bubble layer. The thickness of the sublayer is determined by a force balance on the vapor blanket, which is also related to the condition of the bubble region. About 1100 experimental points have been selected to verify the proposed model. Comparison between the predictions by the proposed model and the experimental result shows a good agreement that more than 90% of these data are predicted within ±20%.

Commentary by Dr. Valentin Fuster
2010;():881-892. doi:10.1115/ICONE18-29920.

According to characteristics of TOPAZ-II reactor, a transient analytic method combined a nuclear reactor six-group point-kinetics model, a reactor core thermal-hydraulic model, a thermionic fuel element (TFE) performance model is established. Afterwards, establishment and debugging of this transient analytic code are completed. Verification results are reasonable agreement with reported American and Russian data. The code is original with the author.

Commentary by Dr. Valentin Fuster
2010;():893-903. doi:10.1115/ICONE18-29933.

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.

Commentary by Dr. Valentin Fuster
2010;():905-913. doi:10.1115/ICONE18-29951.

Plunging jets play an important role in nuclear reactor safety research. In the present paper the case of the strainer clogging issue is considered. Entrained air caused by a plunging jet has an influence of the liquid flow field and on the fibre transport in the sump. In the paper the amount of entrained air is given as an inlet boundary condition according to correlations in the literature and confirmed by own experiments. The influence of entrained air on the fibre deposition pattern at the bottom of a tank and on the mixing procedure for the case of temperature differences between jet and tank water are investigated by CFD calculations and compared to experiments. The presented work is part of a joint research project performed in cooperation between the University of Applied Science Zittau/Görlitz and Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow in Zittau and the development of CFD models for its simulation in Rossendorf (Krepper et al. 2008).

Commentary by Dr. Valentin Fuster
2010;():915-920. doi:10.1115/ICONE18-29953.

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.

Commentary by Dr. Valentin Fuster
2010;():921-932. doi:10.1115/ICONE18-29954.

In recent years, many NPPs have developed and implemented severe accident management guidelines (SAMG). It is the primary objective of developing SAMG to prevent or mitigate the consequences of severe accidents by keeping the reactor pressure vessel (RPV) integrity and reducing the load to the containment. In a hypothetical Station Blackout accident all active safety systems are unavailable. Without additional measures this would lead to heating-up of the reactor core with severe core degradation. To avoid or to limit the consequences of a possible core heat up, different accident management strategies can be applied. This paper presents an assessment of early-phase accident management actions for VVER-1000 reactors. In particular Primary Side Depressurization (PSD) is investigated as a basic strategy for managing severe accidents under high pressure conditions. In addition, Secondary Side Depressurization (SSD) is also being investigated. It aims at fast reduction of the secondary pressure and feeding the steam generators’ secondary side with water from the feed water tank or from a different source. In that way, the heat removal from the primary to the secondary side can be significantly enhanced and the core heat-up at high pressure can be delayed. A number of simulations with different criteria for actuation of the PSD procedure and additional SSD were performed using the thermal-hydraulic system code ATHLET. This paper provides a detailed modelling of the reactor coolant system and the required safety systems, analysis of the thermal-hydraulic and safety parameters and description of the physical phenomena. Special attention is given to the possibilities of preventing or at least delaying an extended core heat-up depending on the availability of the operational and safety systems. The effectiveness of the applied accident management measures and the effect on the accident progression were studied in order to assess the maximum response time for operators’ intervention.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():933-942. doi:10.1115/ICONE18-29968.

The paper reports on the results carried out from the natural circulation and gas-injectio