ASME Conference Presenter Attendance Policy and Archival Proceedings

2014;():V02AT00A001. doi:10.1115/ICONE22-NS2A.

This online compilation of papers from the 2014 22nd International Conference on Nuclear Engineering (ICONE22) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Thermal Hydraulics

2014;():V02AT09A001. doi:10.1115/ICONE22-30009.

An earthquake is one of the most serious phenomena for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically, to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In the case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and a shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.

Therefore, in this research project, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate the effects of vibration on bubbly flow in the components and construct an experimental database for validation, we performed visualization experiments of vertical bubbly flow in a rectangular water tank on which a sine wave vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of bubbles, shapes of bubbles and liquid velocity distributions around bubbles, were shown. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble tilt angle was almost same as the cycle of oscillation table velocity.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A002. doi:10.1115/ICONE22-30015.

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and to use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basis accidents and beyond design basis accidents of HTR10-GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake under the loss of the system pressure. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted. At the same time, the system pressure was supposed to lose by some reason. Thus the reactor power increased and the temperature of the core increased. And the protection system warns with two scram signal: too high of the negative varying rate of the system pressure and too high of the reactor power, which should induce the reactor to scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure continued to increase but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1203.4°C. It was a little bit lower than 1230°C — the fuel temperature limitation of HTR-10 and there is no release of any radioactivity. So the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10.

Topics: Pressure
Commentary by Dr. Valentin Fuster
2014;():V02AT09A003. doi:10.1115/ICONE22-30027.

The flow transient critical heat fluxes (FT-CHFs, qcr,sub) in a SUS304-circular tube caused by a rapid decrease in velocity from non-boiling regime are systematically measured for initial flow velocities (u0=7.057 to 13.635 m/s for conditions of u0=6.9, 9.9 and 13.3 m/s), initial heat fluxes (q0=15.59 to 17.34 MW/m2), inlet liquid temperatures (Tin=290.12 to 308.51 K), outlet pressures (Pout=698.38 to 1288.97 kPa) and decelerations caused by a rapid decrease in velocity (u(t)=u0+αt, α=−7.357 to −0.326 m/s2) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304-circular tubes of inner diameter (d=6 mm), heated length (L=59.5 to 59.7 mm), effective length (Leff=48.7 to 50.2 mm), L/d (=9.92 to 9.95), Leff/d (=8.12 to 8.37) and wall thickness (δ=0.5 mm) with average surface roughness (Ra=3.89 μm) are used in this work. The flow transient CHFs for SUS304-circular tube are compared with authors’ steady-state CHF data for the empty VERTICAL and HORIZONTAL SUS304-circular tubes and the values calculated by authors’ steady-state CHF correlations against outlet and inlet subcoolings for the empty circular tube. The influences of initial flow velocity (u0), initial heat flux (q0) and deceleration caused by a rapid decrease in velocity (α) on the flow transient CHF are investigated into details and the widely and precisely predictable correlations of CHF and flow velocity at the flow transient CHF for the circular tube is given based on the experimental data. The correlations can describe the flow velocity and the CHFs at the flow transient CHFs for SUS304-circular tube obtained in this work within ±20 % difference.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A004. doi:10.1115/ICONE22-30039.

Steam injector (SI) is a passive jet pump with a converging-diverging structure. A SI operates without an electrical power source by direct contact condensation of a supersonic steam flow and a subcooled water jet in a mixing section. Furthermore, there are advantages that a SI has high heat-transfer performance and discharges water at high pressure. Therefore a SI is expected to apply to the safety system that is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant. However it is not cleared about the operating range of a SI which is taken account of the discharged flow structure.

The objective of the present study is to reveal the influence of two-phase flow behavior on operating limits of a SI. The test section of the SI is made by transparent material to observe flow structure in it. The pressure distributions along the flow direction and the discharge pressure were measured by changing the inlet steam pressure and a load on exit of the SI. At the same time, the discharged flow at the diffuser was observed with a high speed camera.

From the observation results, it was confirmed that a boundary which the flow structure changed in the discharged flow. The area from the throat to this boundary is seemed a two-phase region that steam which has not been condensed completely in the mixing nozzle remains. It was found that this boundary moved to upstream as the load on the exit increased and significant pressure rise occurred at the position the boundary reached. Additionally white propagations toward downstream were observed. This propagation is seemed a pressure wave propagation. The velocity of the propagation was estimated by image processing. Assuming a pressure wave propagates at sonic speed, void fraction at the discharged flow was estimated by existing homogenized model.

From the above, the influence of two-phase flow in discharged flow on the operating limits of the SI is discussed.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A005. doi:10.1115/ICONE22-30042.

Water ingress accident is one of the most severe accidents which must be analyzed in high temperature gas-cooled reactor pebble-bed modular (HTR-PM). The droplet could enter the primary circuit under the design basis accident of a double-ended guillotine break of a heat transfer tube. This paper simulates the behavior of single droplet evaporation and movement in the steam generator by numerical methods. Based on the structure characteristics of steam generator, the life time of droplet and the distance that the single droplet could move have been analyzed. The important parameters such as the droplet diameter, helium temperature, helium pressure and helium velocity which have an influence on the behavior of droplet evaporation and movement have also been discussed in detail.

The preliminary numerical simulation results indicate that the droplet diameter, helium velocity and helium temperature play an important role in the life time of droplet in the accident situation in the primary circuit. Helium pressure has a little effect on droplet evaporation in practical situation. The numerical simulation results demonstrate that only certain droplets with a diameter in certain range could arrive to the bottom of the steam generator pressure vessel (SGPV) and enter into the steam generator annular channel after collision with the bottom of the SGPV. The distance that the single droplet could move in the primary circuit is decided by a various complex factors such as the structure of the primary circuit, the droplet diameter and helium velocity. The preliminary analyses indicate that there is little probability for the single droplet to enter into the reactor core of the HTR-PM.

Topics: Drops , Boilers , Evaporation
Commentary by Dr. Valentin Fuster
2014;():V02AT09A006. doi:10.1115/ICONE22-30049.

Water ingress is one of the peculiar accidents of high temperature gas-cooled reactors (HTR). Since the pressure of the secondary circuit is higher than that of the primary circuit, when the break of heating tubes of the SG occurs, liquid water and steam will enter the primary circuit from the secondary side, and then enter the reactor core together with the helium coolant. This will result in introduction of a certain positive reactivity, rising of the primary pressure and corrosion of the in-core fuel elements and graphite components. Therefore, water ingress is a complex physical processes involving water spout, flash evaporation, multi-phase flow, heat transfer and other phenomena, and in-depth research of this accident will benefit further understanding of the characteristics of HTRs.

Based on the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), assuming that the break of steam generator (SG) heating tube can be considered as a round pressure nozzle, the characteristics of the flow line at breaks with different diameters, as well as the property of the droplet spectrum are studied. According to the analysis of droplet properties of flow and evaporation in the primary circuit, the preliminary result shows that under the design basis accidents of water ingress, the water will enters the reactor core in the form of steam. This analysis result can provide a basis for the further research of water ingress of HTR in the future.

Topics: Drops , Accidents , Water
Commentary by Dr. Valentin Fuster
2014;():V02AT09A007. doi:10.1115/ICONE22-30051.

This paper demonstrates the validity of newly-developed the wall drag and form loss partitioning methods for dispersed flow. The wall drag partitioning method is proposed based on the equation of a solid/fluid particle motion. The bubble is faster than the water in a contraction while the former is slower than the latter in an expansion. The droplet is slower than the gas in a contraction while the former is faster than the latter in an expansion. In addition, this study shows that the existing form loss model predicts incorrectly the dispersed phase velocity. A new form loss computation method is proposed.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A008. doi:10.1115/ICONE22-30072.

The fully ceramic microencapsulated (FCM) fuel, one of the accident tolerant fuel (ATF) concepts, consists of TRISO particles randomly dispersed in SiC matrix. This high heterogeneity in compositions leads to difficulty in explicit thermal calculation of such a fuel. For thermal analysis of a fuel element in very high temperature reactors (VHTRs) which has a similar configuration to the FCM fuel, a two-temperature homogenized model was proposed by the authors. The model was developed using particle transport Monte Carlo method for heat conduction problems. It gives more realistic temperature profiles and provides the fuel-kernel and graphite temperatures separately.

In this paper, we apply the two-temperature homogenized model to single-channel thermal analysis of the FCM fuel element for steady- and transient-states using 2-D FEM/1-D FDM hybrid method. The results of analyses are compared to those of conventional UO2 fuel having the same geometric dimension and operating conditions.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A009. doi:10.1115/ICONE22-30083.

The behavior of graphite dust is important to the safety analysis of High-Temperature Gas-cooled Reactor (HTGR). The fission products released by fuel elements would enter the primary loop and combine with dust, resulting in that the dust has a high load capacity of cesium, strontium, iodine and tritium. It would bring difficulty and inconvenience to the maintenance and repair of steam generator. Therefore, the behavior of graphite dust in the steam generator is essential to the safety of High Temperature Gas-cooled Reactors. The present study focused on the deposition and resuspension of graphite dust in steam generator of HTR by numerical method. The results show that the graphite dust in steam generator deposits on the surface of heat transfer tube through turbulent deposition, thermophoretic deposition, and other depositional mechanisms, of which thermophoretic deposition is the main mechanism for the particles with the diameter of 2.2μm in the present study. The preliminary calculation result shows that about 6760mg/m2 of graphite dust tends to load on the tube surface.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A010. doi:10.1115/ICONE22-30090.

Experimental and numerical studies into thermal stratification by direct steam condensation in a torus type suppression pool were carried out to investigate the reactor core isolation cooling in the accidents of Fukushima Daiichi nuclear power plants. The suppression pool was manufactured to be a 1/22 scaled model of a Fukushima Daiichi nuclear power plant. Two different types of spargers were employed to simulate different units of the plants. In a sparger, 132 holes were uniformly drilled on the side of a pipe. However, the other sparger injected steam to the bottom. Flow rate was varied in a wide range to examine the effect on thermal stratification in the suppression pool. The experimental results showed that the sparger type influenced formation of thermal stratification. Moreover, steam flow rate strongly affected the onset time of thermal stratification, and the disappearance of the thermal stratification was affected by subcooling temperature. Computer simulation using a commercial software was conducted and the results show similar temperature profiles to the experimental results. Steam condensation was visualized in a vicinity of the spargers using high speed camera.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A011. doi:10.1115/ICONE22-30091.

In the Microwave Heating de-nitration (MH) method developed in Japan, a mixed solution of uranyl nitrate and plutonium nitrate (Pu/U mixed nitrate solution) recovered from the spent fuel in the reprocessing plant is converted directly to mixed oxide (MH-MOX) powder. This MH-MOX powder is utilized to fabricate MOX fuel with UO2 powder for FBR.

The MH method is accompanied with transient boiling phenomena such as overflow and flushing. Toward high-speed and high-capacity conversion by MH-method in the future, it is required to avoid overflow and flushing and to understand optimal conditions for vessel shape design and microwave output operation. The objective of this paper is to elucidate occurrence criteria of flushing phenomena.

At the first step for this objective, basic knowledge of transient boiling phenomena by the MH-method has been mainly acquired with using distilled water. From the results, it is observed that generation of singular bubble triggers flushing and distilled water just before flushing is superheated more than 10 °C in conditions that flushing is confirmed. Also, the water temperature reaches its peak, and it is almost unchanged in conditions that flushing is not confirmed. In no flushing conditions, it is found that the evaporation starts from the point where the water temperature reaches its peak, and water level is decreased gradually. Thus, the difference of thermal characteristics greatly affects whether flushing occurs.

The second step, by focusing on the process before flushing occurs, we investigate flow structure, and heat amount through the temperature distribution inside the jelly just after microwave heating. Potassium chloride (KCl) aqueous solution and water are used. KCl solution is electrolyte solution same as reprocessing solution. From the results, upward flow is observed near the center of the solution in the case of the water just after microwave heating. On the other hand, downward flow is observed in the case of KCl solution. As a result of temperature distribution, the water is heated its near center, and KCl solution is heated the around. Thus, it is considered that heat amount of microwave varies depending on solution characteristics, and the difference greatly affects the flow structure. The heat amount and flow structure are critical factors that can determine transient boiling phenomena. We need to investigate transient boiling phenomena of KCl solution in the future.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A012. doi:10.1115/ICONE22-30098.

This paper presents a set of numerical procedure to innovate CFD code into a PWR component analysis code. A porous media approach is adapted to two-fluid model and conductor model, and a pack of constitutive relations close the numerical model into a PWR component analysis code. The separate verification calculations on conductor model and porous media approach, and the validation calculation for the integrated component-scale code are introduced.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A013. doi:10.1115/ICONE22-30120.

Genetic algorithm (GA) has been widely applied in optimal design of nuclear power components. Simple genetic algorithm (SGA) has the defects of poor convergence accuracy and easily falling into the local optimum when dealing with nonlinear constraint optimization problem. To overcome these defects, an improved genetic algorithm named dual-adaptive niched genetic algorithm (DANGA) is designed in this work. The new algorithm adopts niche technique to enhance global search ability, which utilizes a sharing function to maintain population diversity. Dual-adaptation technique is developed to improve the global and local search capability at the same time. Furthermore, a new reconstitution operator is applied to the DANGA to handle the constraint conditions, which can avoid the difficulty of selecting punishment parameter when using the penalty function method. The performance of new algorithm is evaluated by optimizing the benchmark function. The volume optimization of the Qinshan I steam generator and the weight optimization of Qinshan I condenser, taking thermal-hydraulic and geometric constraints into consideration, is carried out by adopting the DANGA. The result of benchmark function test shows that the new algorithm is more effective than some traditional genetic algorithms. The optimization design shows obvious validity and can provide guidance for real engineering design.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A014. doi:10.1115/ICONE22-30153.

An earthquake is one of the most serious phenomena to consider for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors were contracting with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown in seismic conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and shear force through the pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation during the earthquake.

Therefore, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in a series of studies. In this study, to develop the predictive technology of two-phase flow dynamics under earthquake acceleration, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT (Two-Phase Flow simulation code with Interface Tracking) was expanded to two-phase flow simulation in seismic conditions. In a previous study, we performed a numerical simulation of a two-phase bubbly flow in a horizontal pipe and a vertical bubble motion in a water tank in seismic conditions. And it was confirmed that the modified TPFIT can be applicable to the bubbly flow in seismic conditions.

In this paper, the two-phase bubbly flow in a simulated single-subchannel excited by oscillation acceleration was simulated by using the expanded TPFIT. A calculation domain used in this simulation was a simplified subchannel in a BWR core. And time-series of void fraction distributions were evaluated based on predicted bubble distributions. When no oscillation acceleration was added, void fraction concentrated in a region near the wall. When oscillation acceleration was added, void fraction distribution was changed by time. And coalesces of bubbles occurred in the numerical simulation, and bubbles with relatively large diameter were observed. In the results, complicated void fraction distribution was observed, because the response of void fraction distribution on the oscillation acceleration was dependent on not only imposed acceleration, but also the bubble diameter.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A015. doi:10.1115/ICONE22-30161.

A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of a void fraction in a 5×5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied in a boiling two-phase flow experiment under the atmospheric conditions assumed in an accident and spent fuel pool. The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. Wire electrodes 0.2 mm in diameter are arranged in latticed patterns between the rod bundle, while a conductance value in a region near one wire and another gives a local void fraction in the central-subchannel region. 32 points (= 6×6−4) of the local void fraction can be obtained as a cross-sectional distribution. In addition, a local void fraction near the rod surface can be estimated by a conductance value in a region near one wire and one rod using the simulated fuel rods as rod electrodes, which allows 100 additional points (=4×25) of the local void fraction to be acquired. The devised sensors are installed at five height levels to acquire two-phase flow dynamics in an axial direction. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The time resolution of void measurement exceeds 800 frames (cross-sections) per second. The heated rod bundle has an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various experimental conditions, such as inlet-flow velocity, rod-bundle power and inlet subcooling. The experimental results exhibited axial and radial distribution of two-phase flow structures, i.e. void-fraction and phasic-velocity distributions quantitatively.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A016. doi:10.1115/ICONE22-30171.

3-D simulation of supercritical water flow instability in parallel channels and a natural circulation loop are presented. Results are obtained for various heating powers. The results show that, in the natural circulation loop the steady state mass flow will firstly increase with the heating power and then decrease. And mass flow grows with the growing of the inlet temperature, decreases with the growing of system pressure. Under a large heat flux, the parallel channels will experience the flow instability of out phase mass flow oscillation. And the oscillation amplitude will grow with the growing of heating power. At last, the numerical simulations are validated by B.T. Swapnalee’s experience formula.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A017. doi:10.1115/ICONE22-30190.

This paper presents a CFD study of the flow and deposition of particles in vapor in a 1000mm by 20mm narrow rectangular channel. The temperature of the wall is set below the condensation point, so the vapor will change into the liquid phase near the wall. The flow and deposition of the particles in the condition when the phase transition of water occurs is simulated by ANSYS Fluent code. The result shows that the condensation of the water vapor will change the temperature field and the velocity of the flow in the channel. The particles will be forced to the wall due to the thermophoretic force. The condensation developing process will enhance the deposition of the particles while the developed liquid layer will stop the particles flow to the wall.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A018. doi:10.1115/ICONE22-30198.

The conceptual design and basic engineering of a multipurpose ADS for R&D, named Chinese Initiative Accelerator Driven System (CIADS), have been carried out by Chinese Academy of Science. A lead-bismuth eutectic (LBE) spallation target with beam window is an alternative for the CIADS. In this work, by using the code system ANSYS, the static structure analysis of the spallation target beam window at the initial state was performed parametrically in conjunction with the thermal hydraulics to check the design compliance with the stress and the buckling design criteria. The static structure analysis of the target window was presented under the conditions of different window thicknesses, proton beam intensities and beam spot diameters. The results showed that all the parameters investigated in this work which meet the needs of thermal hydraulics design also compliance with the stress and the buckling design criteria. Also, the results were used to select the thickness of the beam window to maximize the engineering margins. More work will be done in the future for the structural analysis of target operating with certain dosage irradiated by the neutrons and the protons.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A019. doi:10.1115/ICONE22-30199.

Several studies have proposed the use of nanofluids to enhance the in-vessel retention (IVR) capability in the severe accident management strategy implemented at certain light-water reactors. Systems using nanofluids for IVR must be applicable to large-scale systems, i.e., infinite heated surfaces. However, the effect of the size of heater with nanoparticle deposition was revealed that the CHF is decreased with the increased heater size. On the other hand, the CHF using a honeycomb porous plate was shown experimentally to be more than twice that of a plain surface with a heated surface diameter of 30 mm, which is comparatively large compared to 10 mm. This enhancement is resulted from the capillary supply of liquid onto the heated surface and the release of vapor generated through the channels.

In the present paper, in order to enhance the CHF of a large heated surface, the effects of a honeycomb porous plate and a nanofluid on the CHF were investigated experimentally. As a result, the CHF was enhanced greatly by the attachment of a honeycomb porous plate to the modified heated surface by nanoparticle deposition, even in the case of a large heated surface.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A020. doi:10.1115/ICONE22-30200.

Within the framework of large-scale MOCKA (KIT, Germany) experiments, a series of experiments have been performed to study the interaction of a simulant oxide (Al2O3, ZrO2, CaO) and metal melt (Fe) in a stratified configuration. To allow for a longer-term interaction, additional heating was provided by alternating additions of thermite and Zr metal to the melt. Since the heat generated by the thermite reaction and the exothermal oxidation reaction of Zr is mainly deposited in the oxide phase, prototypic heating of both melt phases is achieved. This allows the investigation of concrete erosion by metal melt as well as by the oxide which was not possible in all former experiments.

Current tests in the MOCKA (KIT, Germany) program are focused on assessing the influence of concrete reinforcement (rebars) on the cavity erosion behaviour using a simulant oxide-iron melt in a stratified configuration. The experiments are performed in siliceous concrete crucibles with an inner diameter of 25 cm containing 12 wt.% reinforcement. In these experiments, the overall downward erosion by the metal melt was of the same order as the sideward one. In addition, the lateral erosion in the overlaid oxide melt region was about the same as in the metal melt region. The former experiments (BETA, COMET-L) and MOCKA tests on siliceous concrete without reinforcement have produced results with pronounced downward erosion by the metal phase. This pronounced downward erosion of the siliceous concrete without rebars seems to be inherent for melts containing a significant fraction of iron.

Topics: Metals , Concretes
Commentary by Dr. Valentin Fuster
2014;():V02AT09A021. doi:10.1115/ICONE22-30205.

An experiment on a PWR station blackout transient with the TMLB’ scenario and accident management (AM) measures was conducted using the ROSA/large scale test facility (LSTF) at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator (ACC) tanks. The AM measures proposed in this study are steam generator (SG) secondary-side depressurization by fully opening the safety valves in both SGs with the start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed the primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after the completion of ACC coolant injection. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of SG U-tube collapsed liquid level and primary mass flow rate after the gas ingress. The SG coolant injection flow rate was found to affect significantly the peak cladding temperature and the ACC actuation time through the RELAP5 sensitivity analyses.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A022. doi:10.1115/ICONE22-30207.

In Pb Bi loop, the turbulent buffeting phenomenon of lead bismuth fluid in heat exchanger, may cause fatigue damage of heat transfer tube. Through the establishment of Pb-Bi loop heat exchanger model, invent a program for calculation, we can get the turbulent buffeting characteristics. The results show that: As the increases of heat exchange tube cross number, the buffeting coefficients pipeline is greater, the probability of occurrence of turbulent buffeting phenomenon is more low; The greater lead bismuth flow velocity is, the more prone to turbulent buffeting phenomenon. With vertical heat tube center distance increases, the buffeting coefficients decreased first and then increased, when the buffeting coefficients reached the minimum value, the turbulent buffeting phenomenon is most intense; As the horizontal heat pipe center distance is bigger, turbulent buffeting phenomenon is becoming less clear.

Topics: Turbulence
Commentary by Dr. Valentin Fuster
2014;():V02AT09A023. doi:10.1115/ICONE22-30214.

Liquid metals have been used as coolants of several kinds of nuclear reactors, and the prediction of critical heat flux (CHF) is rather important for the design, safety and economy of these reactors. A film dryout model considering the deposition and entrainment of droplets was established to obtain the CHF of liquid metal in annular flow flowing in tubes. The correlations of deposition rate, entrainment rate and so on for conventional fluids were used, and the initial entrainment fraction was determined according to experimental data. Results showed that the correlations for conventional fluids could be used for liquid metals approximately, but relatively large error might occur for large heat flux. The accuracy of this model for sodium and potassium was similar for small heat flux, but had some differences for large heat flux. Special correlations of deposition rate, entrainment rate and so on should be developed to predict the CHF of liquid metals more accurately.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A024. doi:10.1115/ICONE22-30237.

Simulations of complex scenarios in nuclear power plants have been improved by the utilization of coupled thermal hydraulic (TH) and neutron kinetics (NK) system codes with the development of computer technology and new calculation methodology which made it possible to perform transport calculation schemes with accurate solutions. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0.0 code. By using this code, a multi-dimensional neutron kinetics model based on the NESTLE code can be implemented also. In this way, during a 3D TH/NK coupled simulation, RELAP5-3D calls the appropriate NESTLE subroutines to perform the calculations. The development and the assessment of the thermal hydraulic RELAP5 code model for the IPR-R1 TRIGA have been validated for steady state and transient situations and the results were published in preceding works. The model has been adapted to RELAP5-3D code and was verified to point kinetic calculations. After this, adequate cross sections to the NK code were supplied using the WIMSD5 code. The results of steady state and transient calculations using the 3D neutron modeling to the IPR-R1 are being presented in this paper.

Topics: Neutrons
Commentary by Dr. Valentin Fuster
2014;():V02AT09A025. doi:10.1115/ICONE22-30244.

Thermal hydraulic characteristics of liquid sodium flowing in an annulus are experimentally studied. The annulus is 1100 mm in length, 6 mm as inside diameter and 10 mm as outside diameter. The heat flux in the experiment is from 50 to 210 kW/m2, with Re number from 0 to 18000 and average fluid temperature from 200 °C to 500 °C. Experimental data show that the flow regime of liquid sodium flowing in the annulus can be divided into three regions including laminar flow (Re<2000), transition flow (2000<Re<4000) and turbulent flow (Re>4000). The effects of heat flux, Re number and average fluid temperature of the test section on the heat transfer coefficient are investigated separately. For different regions, correlations for the friction coefficient and for the Nu number are obtained from the experimental data.

Topics: Annulus , Sodium
Commentary by Dr. Valentin Fuster
2014;():V02AT09A026. doi:10.1115/ICONE22-30245.

Experimental and theoretical studies have been performed for entrainment at T-junction consisting of hot leg and ADS-4 pipe. Research of basic influence factors on onset entrainment and stable entrainment is conducted and the effects of gas chamber height and gas Froude number on entrainment are discussed on the basis of experimental data. Based on visualization research, the flow regime in the main pipe is recorded by high-speed camera. Three kinds of entrainment phenomena including stratified entrainment, annular entrainment and intermittent entrainment can be clearly found, and feature of each phenomenon is summarized. Also, Onset entrainment model is established and accords with experimental data well.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A027. doi:10.1115/ICONE22-30250.

One of the key technologies for the development of the PMR200, a VHTR demonstration plant, is a verification of the reactor cavity cooling system (RCCS) performance, which ensures reactor safety by passively removing heat from the reactor cavity. A preliminary numerical analysis of the RCCS showed that the maximum temperature in RCCS reached up to 700°C. Since radiation dominates the heat transfer at such a high temperature, it should be considered in both the design and associated numerical works for the test facility. For a verification of the RCCS performance, a 1/4 scale test facility has been constructed, and a performance test is being carried out. As the first step for the design of the test facility, a scaling analysis has been performed; and the ratio of variables between the model and prototype were determined. Numerical calculations using a CFD code were also performed to support the scaling analysis. It was confirmed that the scaling analysis was reasonably correct.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A028. doi:10.1115/ICONE22-30256.

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. Currently, steam generators operate under a leak-before-break approach. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. Choked flow of subcooled water through small cracks such as in steam generator tube wall cracks is studied both with experiments and analytical models. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. Slits of very small channel length to hydraulics diameter ratio (L/D) were manufactured and tested upto 6.89 MPa pressure and range of subcoolings 10–40 °C. Small flow channel length was used (1.3mm) equivalent to steam generator tube thickness with differences in surface roughness. The effect of L/D on the choking flow rates was examined and was contrasted with other data in literature. Analytical models were applied highlighting the importance of non-equilibrium effects and the effects of L/D ranging from 1.3 to 400 on the chocked flow were investigated.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A029. doi:10.1115/ICONE22-30259.

The condensation heat transfer occurring in containment atmospheres during the loss of coolant accident (LOCA), is one of the most important areas in research related to the safety of nuclear reactors. In the advanced Generation III and III+ nuclear reactors, decay heat is removed by passive containment cooling system (PCCS). For the system, the study of condensation of steam in the presence of non-condensable gases is prior to be investigated because when LOCA happens steam flashes into the containment which contains air and other non-condensable gases (helium, argon, etc.). An experimental investigation has been conducted to evaluate the steam heat removal capacity over a vertical tube external surface with air. Condensation heat transfer coefficients have been obtained under the total pressure ranging from 0.4MPa to 0.6MPa, the wall subcooling ranging from 13 to 25°C and air mass fraction ranging from 0.07 to 0.52. The influence of the wall subcooling on the steam condensation heat transfer with the fixed pressure and air mass fraction have been researched. The effect of wall subcooling on condensation heat transfer coefficient with air is negative. The developed empirical correlation for the heat transfer coefficient covered all data points within 15%.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A030. doi:10.1115/ICONE22-30260.

The characteristics of two-phase slug flow in a narrow rectangular channel with cross section of 3.25 mm × 43 mm under vertical and inclined conditions are investigated using a high speed video camera system. It is found that the velocity of Taylor bubble in vertical continuous slug flow could be well predicted by the Nicklin et al. (1962) correlation, in which C0 is given by the correlation of Ishii (1977), and the drift velocity given by the correlation of Sadatomi et al. (1982) or Clanet et al. (2004). For low two-phase superficial velocity (FrTP ≤ 3.5), the Taylor bubble velocities gradually increase with the increasing in inclination angles and almost approximate the maximum value for θ = 30°. For high two-phase superficial velocity (FrTP > 3.5), the influence of the inclination angles on the Taylor bubble velocity is insignificant, and the bubble velocity under vertical condition is slight lower than those under inclined conditions. For the inclined cases, the nose of Taylor bubble is deviated from the centerline and its position is the function of the two-phase superficial velocity as well as the inclination angle.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A031. doi:10.1115/ICONE22-30265.

In order to gain more insights into the system depressurization and entrainment behavior after actuation of the fourth-stage (ADS-4) valves during a loss-of-coolant-accident, the ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) scaled to AP1000 was constructed to simulate the accident scenario with air-water and steam-water. A brief scaling analysis with emphasis on related thermal hydraulic processes was presented. Entrainment phenomena at vertical up tee branch were observed and analyzed. Preliminary test data of onset of entrainment and entrainment rate were collected with air-water tests and relevant conclusions were obtained.

Topics: Coolants , Accidents , Valves , Steam , Water
Commentary by Dr. Valentin Fuster
2014;():V02AT09A032. doi:10.1115/ICONE22-30271.

Owing to its compact structure, the once-through steam generator (OTSG) is used widely in integrated Pressurized Water Reactor (PWR) The casing OTSG in concentric annuli tube is a new type of steam generator which applies double sides to transfer heat. The water of the secondary side goes through complicated phase change processes, and the flow pattern and heat transfer cases are much more complex than those of the natural circulation steam generator used in PWR. It is necessary to study their steady-state and dynamic characteristics. By means of THEATRe code which was based on the two-phase drift flux model and was modified by adding module calculating the effect of rolling motion, the casing OTSG simulation model in rolling motion was built. It can describe the parameters change in every section of OTSG accurately in rolling motion, and can embody dynamics characteristics from different aspects. By comparing the operational data in different rolling amplitude and rolling period, flow operational characteristics and principles were analyzed. The results can be used to analyze the thermal-hydraulics characteristics of the integrated PWR in rolling motion.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A033. doi:10.1115/ICONE22-30281.

The sodium-cooled fast reactor (SFR) severe accident analysis computer code SIMMER-III has been developed and assessed comprehensively and systematically in a code assessment (verification and validation) program which consists of a two-step effort: Phase 1 for fundamental or separate-effect assessment of individual code models; and Phase 2 for integral assessment of key physical phenomena relevant to SFR safety. This paper describes the achievement of the code assessment on material expansion dynamics in the framework of the Phase 2 assessment program. Detailed descriptions are given for two representative experimental analyses (VECTORS and OMEGA), which are intended to validate high speed multi-phase flow dynamics in pin bundle structure and large vapor bubble expansion dynamics into a coolant pool, respectively. Through the assessment program, the SIMMER-III code has proved to be basically valid both numerically and physically, with current applicability to integral reactor safety calculations.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A034. doi:10.1115/ICONE22-30282.

An thermo-hydraulic analysis model for the reflood phase was established in the paper, based on the flow and heat transfer characteristics of the reflood. A code based on the model was developed for the thermo-hydraulic analysis of reflood. By comparing and analyzing the calculation results with the experimental results, the influences of the system pressure and the subcooling of coolant on reflood phase were studied. The study also provided theoretical basis for safety analysis of fuel element of reflood phase.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A035. doi:10.1115/ICONE22-30283.

The quenching characteristics of particulate debris bed during bottom and top flood is analyzed in this paper. The top flood model is formulated by dividing the quenching process into downward frontal period and upward frontal period, which are controlled by the counter-current flow limitation (CCFL) condition and effects of the incoming coolant subcooling and steam cooling in dry channels during quenching process. The bottom flood model is based on porous media theory under the assumption that the height of the two phase region is negligible and the particulate debris bed is divided into single phase liquid and single phase vapor region. The results calculated by these models were compared with the experimental data. The influences of porosity, initial debris temperature and other parameters on both the top and bottom quenching process were studied in this paper. During the top flood, the quenching velocity increased with the increase of the porosity and the decrease of the initial debris temperature. The porosity and initial debris temperature had a larger influence on quenching velocity compared with other parameters, such as initial coolant temperature and coolant flow rate. During the bottom flood, the quenching velocity also increased with the increase of the porosity and the decrease of the initial debris temperature. However, the coolant flow rate had a large influence on the quenching velocity unlike that during the top flood. Quenching from bottom may be superior to the quenching from top. The results can be expected to be useful to evaluate the quenching process of the particulate debris bed.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A036. doi:10.1115/ICONE22-30295.

Experimental study on resistance of air-water two-phase flow in a vertical 3 × 3 rod bundle was carried out under normal temperature and pressure. The rod diameter and pitch were 8 mm and 11 mm, respectively. The ranges of gas and liquid superficial velocity were 0.013∼3.763 m/s and 0.076∼1.792 m/s, respectively. The result indicated that the existing correlations for calculating frictional coefficient in the rod bundle and local resistance coefficient could not give favorable predictions on the single-phase experimental data. For the case of two-phase flow, eight correlations for calculating two-phase equivalent viscosity poorly predicted the frictional pressure drop, with the mean absolute errors around 60%. Meanwhile, the eight classical two-phase viscosity formulae were evaluated against the local pressure drop at spacer grid. It is shown that Dukler model predicted the experimental data well in the range of Rel<9000 while McAdams correlation was the best for Rel⩾9000. For all the experimental data, Dukler model provided the best prediction with MRE of 29.03%. Furthermore, approaches to calculate two-phase frictional pressure drop and local resistance were proposed by considering mass quality, two-phase Reynolds number and densities in homogenous flow model, resulting in a good agreement with the experimental data.

Topics: Fuel rods
Commentary by Dr. Valentin Fuster
2014;():V02AT09A037. doi:10.1115/ICONE22-30296.

In 2012 on the large-scale thermalphysic PSB-VVER facility modelling of three different type accidents is executed:

- Guillotine rupture of the pipeline on a reactor inlet;

- A small leak from “the cold” pipeline;

- Rupture of pressurizer surge line.

In experiments work of passive safety systems of new projects of NPP with RU VVER in the conditions of loss of all sources of an alternating current was modelled.

The purpose of experiments was research of influence of new passive safety systems on a temperature condition of a fuel cladding. On the basis of the received experimental data verification of the Russian system codes TRAP-KS, KORSAR/GP and SOCRAT is executed.

In paper the basic results of researches are presented.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A038. doi:10.1115/ICONE22-30297.

This paper presents some consideration and analysis on the flow instability inside the tubes of the moisture separator reheater (MSR), which is a combined equipment of the mist separator and the in-tube side condensation multi-tube type heat exchanger for superheating the exhaust from the high pressure turbine to the low pressure turbine. The condensed water in the tubes will be exhausted by the friction of excess steam vented through the tubes, and with such design, the flow in the tubes will be in an annular gas-liquid two-phase regime, ensuring the steady drainage of condensed water without subcool and the associated temperature oscillation at the tube end. The mechanism of the unstable two-phase flow dynamics in the tubes is discussed. The analysis and the results presented in this paper will be of interests to the related researchers and engineering applications.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A039. doi:10.1115/ICONE22-30301.

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A040. doi:10.1115/ICONE22-30313.

An experiment has recently been performed at Xi’an Jiaotong University to study the wall temperature and pressure drop at supercritical pressures with upward flow of water inside a 2×2 rod bundle. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. Experimental parameters covered the pressure of 23–28 MPa, mass flux of 350–1000 kg/m2s and heat flux on the rod surface of 200–1000 kW/m2. According to the experimental data, it was found that the circumferential wall temperature distribution of a heated rod is not uniform. The temperature difference between the maximum and the minimum varies with heat flux and/or mass flux. Heat transfer characteristics of supercritical water in bundle were discussed with respect to various heat fluxes. The effect of heat flux on heat transfer in rod bundles is similar with that in tubes or annuli. In addition, flow resistance reflected in the form of pressure loss has also been studied. Experimental results showed that the total pressure drop increases with bulk enthalpy and mass flux. Four heat transfer correlations developed for supercritical pressures water were compared with the present test data. Predictions of Jackson correlation agrees closely with the experimental data.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A041. doi:10.1115/ICONE22-30314.

This paper describes the basic design features of the EU-APR1400 reactor core catcher cooling system and its test facility, and the associated scaling analysis model. An assessment of the validity of the scaling analysis using the preliminary performance test result of the test facility is described. This includes comparison of the predicted mass flow rate of the test loop as a function of the heat load to the facility, inlet flow subcooling and system pressure to the experimental results.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A042. doi:10.1115/ICONE22-30315.

According to the literature survey, several scaling studies have been performed to derive a set of scaling criteria which were thought to be suitable for reproducing the major thermal-hydraulic phenomena in a scaled-down CANDU moderator tank similar to that in a prototype power plant during a full power steady state condition [1,2,3]. The objective of building this scaled-down moderator tank is to generate the experimental data necessary to validate the computer codes which are used to analyze the accident analysis of CANDU-6 plants. The major variables of interests in this paper are moderator flow velocity and temperature of the moderator which is D2O inside the moderator tank during a steady state and transient conditions. The reason is that the local subcooling of the moderator is found to be a critical parameter determining whether the stable film boiling can sustain on the outer surface of the calandria tube if the contact of overheated pressure tube and cold calandria tube should occur due to pressure tube ballooning during LBLOCA with ECC injection failure[4]. The key phenomena involved include the inlet jet development and impingement, buoyancy force driven by the moderator temperature gradient caused by non-uniform direct heating of the moderator, and the pressure drop due to viscous friction of the flow across the calandria tube array. In this paper, the previous researches are reviewed, some concerns or potential problems associated with them implied by comparing CFD analyses results between the CANDU-6 moderator tank and 1/4 scaled-down test facility are described, and as a way to examine the assumption of the scaling analysis is true an order-of-magnitude analyses are performed. Based on the results of these analyses the assumption of neglecting (∇*)2V* .and (∇*)2T* terms cannot be justified for the power of 0.5 MW and 1.566 MW for the 1/4 scaled-down facility. Further investigation is thought to be necessary to confirm this result, i.e. if the scaling of the previous work1 is justifiable by some other independent analyses.

Topics: Test facilities
Commentary by Dr. Valentin Fuster
2014;():V02AT09A043. doi:10.1115/ICONE22-30327.

In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17×17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12×12 and 13×13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13×13 and 12×12 arrays) are several times larger than that of the referenced solid fuel (17×17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A044. doi:10.1115/ICONE22-30336.

The characteristics of gas methyl iodide removed by containment filter venting system is similar with bubble column reactor, this study is focused on the absorption performance of methyl iodide in a bubble column reactor with experiment and mathematical calculations, the alkalescent sodium thiosulphate solution in a scale-up facility is used as absorber in the research. The results show that the gas methyl iodide removal efficiency is mainly influenced by the temperature of solution, system pressure, gas flow rate and liquid level respectively. In the range of 0∼80°C, the removal efficiency improve obviously with increasing temperature, while the chemical reaction process is a major factor that limiting the removal efficiency, when the temperature is higher than 80°C, the efficiency is no longer sensitive to the variation of temperature and the mass transfer process become the main limiting factors. The increase of system pressure and height of solution can enhance gas absorption process significantly, the removal efficiency improve linearly with the two parameters. However, the gas volume flow rate plays an opposite role on absorption process mainly because of the reducing on contacting time. In addition, the variation of entrance concentration has a little impact on removal efficiency. The mathematical calculations of removal efficiency fit well with experimental results in the bubble column reactor.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A045. doi:10.1115/ICONE22-30343.

Density Wave Oscillation (DWO) in tubes was usually studied by using the frequency domain method. However, in the conventional model, the heat storage of wall metal was usually neglected to simplify the complex solving process of transfer functions, which might cause unreasonable results when the tube wall had a thick wall or complex geometry structures. Hence, in the present paper, an improved mathematical model was proposed based on the frequency domain theory to theoretically study the DWO in tubes. The present model was an improvement of the conventional model. The most notable improvement in the present model was that the heat storage of the tube wall metal, the internal wall heat flux and the external wall heat flux were all considered as dynamic parameters. Based on the improvement, the prediction of the DWO in tubes by using the present model might be more accurate and reasonable than that by using the conventional model, and this was proved by the comparison of the results obtained with the two models to the experimental results gained from literature. In the present study, it was shown that both the present model and the conventional model could predict the DWO in tubes well when the tube wall was thin, and it was also found that the present model was more appropriate than the conventional model when the tube wall was thick. Both the thickness of the tube wall and the specific heat of tube wall metal play negative roles in the system stability.

Topics: Density , Oscillations , Waves
Commentary by Dr. Valentin Fuster
2014;():V02AT09A046. doi:10.1115/ICONE22-30348.

This study is to investigate the effect of one-dimensional natural circulation on the mixing process of two component gases by evaluating the onset time of natural circulation through the apparatus under the stable density stratified fluid layer.

The experimental apparatus consists of a reverse U-shaped vertical slot and a storage tank. The left side vertical slot consists of the heated wall and the cooled wall. The right side vertical slot consists of the two cooled walls. Temperature difference between the vertical walls was set to 50, 70, and 100 K. In this study, the combination of the two component gases is He/Ar and density ratio of each component is 1.4/10.

The heavy gas was filled with the storage tank and light gas was filled with the reverse U-shaped vertical slot. Before the experiment starts, the localized natural convection was generated in the heated side vertical slot. After the experiment starts, the heavy gas will be transported to the slot by the molecular diffusion and natural convection. And then, natural circulation occurs abruptly through the reverse U-shaped passage. The mixing process of two component gases and the onset time of natural circulation in the vertical fluid layer were affected not only by the localized natural convection but also by the molecular diffusion.

The wall and gas temperatures were measured by thermocouples and the velocity of natural convection was measured to evaluate the characteristics of the mixing process and the natural convection.

These experimental results show that generation time of natural circulation was affected by molecular diffusion and localized natural convection. When the two components of gases have large density ratios and large Gr numbers, the mixing process of two components of gases was affected by more intensively molecular diffusion than localized natural convection when temperature difference was 50K. The mixing process of two component gas was affected by more intensively localized natural convection than molecular diffusion when temperature difference was 70 to 100K. However, two component gases were affected by more intensively molecular diffusion than localized natural convection at small density ratios and small Gr numbers.

Topics: Gases
Commentary by Dr. Valentin Fuster
2014;():V02AT09A047. doi:10.1115/ICONE22-30366.

In this paper, experimental flow and heat transfer data of supercritical pressure HCFC22 flowing in a uniformly heated smooth tube with inner diameter of 1.004 mm at p/pc=1.1 obtained by the authors are analyzed accounting for the influence of the thermophysical properties variation, the buoyancy effect, as well as the flow acceleration effect due to thermal expansion. These analyses indicate that both of the sharp thermophysical properties variation in the fluid adjacent to the wall with low density, low specific heat and low thermal conductivity and the flow acceleration effect due to thermal expansion have significant negative effects on the heat transfer under the present study conditions for HCFC22, while for the friction factor, the thermophysical properties variation is the predominant factor. The buoyancy effect on the flow and heat transfer is negligible.

A new semi-empirical local heat transfer correlation accounting for the thermophysical properties variation and the flow acceleration effect due to thermal expansion for supercritical pressure fluids flowing through a vertical small tube during heating is proposed. The predicted values agree with 95% of the measured data within ±25%. In addition, a flow correlation with thermophysical properties variation correction terms to predict the friction factors for supercritical pressure fluids is proposed which predicts the measured friction factors within ±25%.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A048. doi:10.1115/ICONE22-30378.

In design of the Japan Sodium-cooled Fast Reactor (JSFR), mean velocity of the coolant is approximately 9 m/s in the primary hot leg (H/L) piping which diameter is 1.27 m. The Reynolds number in the H/L piping reaches 4.2×107. Moreover, a short-elbow which has Rc/D = 1.0 (Rc: Curvature radius, D: Pipe diameter) is used in the hot leg piping in order to achieve compact plant layout and reduce plant construction cost. In the H/L piping, flow-induced vibration (FIV) is concerned due to excitation force which is caused by pressure fluctuation on the wall closely related with the velocity fluctuation in the short-elbow. In the previous study, relation between the flow separation and the pressure fluctuations in the short-elbow was revealed under the specific inlet condition with flat distribution of time-averaged axial velocity and relatively weak velocity fluctuation intensity in the pipe. However, the inlet velocity condition of the H/L in a reactor may have ununiformed profile with highly turbulent due to the complex geometry in reactor vessel (R/V). In this study, the influence of the inlet velocity condition on unsteady characteristics of velocity in the short-elbow was studied. Although the flow around the inlet of the H/L in R/V could not simulate completely, inlet velocity conditions were controlled by installing the perforated plate with plugging the flow-holes appropriately. Then expected flow patterns were made at 2D upstream position from the elbow inlet in the experiments. It was revealed that the inlet velocity profiles affected circumferential secondary flow and the secondary flows affected an area of flow separation at the elbow, by local velocity measurement by the PIV (particle image velocimetry). And it was found that the low frequent turbulence in the upstream piping remained downstream of the elbow though their intensity was attenuated.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A049. doi:10.1115/ICONE22-30380.

Passive systems like natural circulation (NC) loops can offer reliable and cost efficient alternatives to common active systems for decay heat removal in nuclear power plants. During the transition between stable single and stable two phase flows, instabilities e. g. flashing and geysering may occur in the riser due to low system pressure and saturation temperature conditions. These instabilities may cause severe stress to the system components. This paper presented some results of the study on the decay heat removal system based on natural circulation, performed on the open loop NC test facility GENEVA, built at TU Dresden in 2013. 16 probes were used to determine void fraction along the riser on nine different levels in high time and spatial resolution, and stability maps was created for riser with inner diameters of 20 mm and 38 mm and up to 85 kW evaporator power.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A050. doi:10.1115/ICONE22-30392.

In spite of most previous studies since 1970, the theory of pulsating pipe flows supported by experimental investigations has not yet completed in comparison with the well-defined theory of steady pipe flows. Therefore, it seems that there is much to be done about experimental research in this field. In order to determine the resistance characteristics of two-phase flow under pulsatile conditions, an experimental investigation on two-phase flow with periodically fluctuating flow rates in a narrow rectangular channel is carried out. A frequency inverter is used to obtain experimental conditions with different fluctuating frequencies, amplitudes and mean values of water mass flow rate. After obtaining experimental results, comparisons between experimental frictional pressure drop values and theoretical calculations have been done. Two-phase flow on pulsating conditions is far more complicated than that on steady conditions because pulsating flow is composed of two parts: a steady component and a superimposed periodical time varying component called oscillation. In this paper, the influence of different fluctuating frequencies, amplitudes and mean values of liquid and gas mass flow rate on two-phase flow pressure drop characteristics is also discussed.

The results show that the total pressure drop and water mass flow rate change with the same fluctuating period except for a phase difference. The phase lag also changes with the fluctuating frequencies and amplitude. The accelerating pressure drop changes dramatically in a fluctuating period, especially at the end of acceleration. Also, the time when the acceleration pressure drop has its maximum value lags the time when the acceleration reaches its peak, mainly because of the inertial of the fluid.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A051. doi:10.1115/ICONE22-30405.

The Modular Accident Analysis Program (MAAP) is a computer code that is used for integrated severe accident analysis. The latest MAAP5 version was validated against the PHEBUS-PF FPT0 and FPT1 tests performed at CEA/IPSN, PBF-SFD Test 1–4 performed at INEL, and QUENCH Test 06 performed at FZK. PHEBUS FPT0, PHEBUS FPT1, and PBF-SFD Test 1–4 are in-pile experiments where a test bundle was housed in the center of a reactor. Comparisons were focused on fuel and shroud temperature histories, and hydrogen generation histories. For the PHEBUS tests, primary system and containment responses were also compared. In general, fuel and shroud temperatures are well predicted by MAAP5. Overall hydrogen mass generated is also well predicted except that MAAP5 over-predicts the total hydrogen mass generated for the FPT1 test. The hydrogen generation at the time of the peak oxidation phase for the FPT0 test is under-predicted while the hydrogen generation for the FPT1 test is over-predicted. In general, MAAP tends to over-predict mass relocations from the upper part of the bundle due to fuel rod collapses by the end of separate effects test transients.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A052. doi:10.1115/ICONE22-30410.

Initiating events such as primary to secondary loss of coolant (PRISE) can lead to conditions forming reversed flow from the second to the primary circuit. Current issue shows the results of a CFD analysis of the distribution of boric acid on the entrance of the core in case of such reversed flow of coolant as a result of PRISE initiation event. Analyzed accident is included in the list of design basis accidents and requires precise approach in analyzing the phenomena associated with the possibility of injection of coolant with low concentration of boric acid in the primary side. The paper emphasizes on the application of CFD to solve the problem. Analyzing the accident is done in advance with the help of system code RELAP. The input data as flow rate, concentration and temperature at the inlet of the reactor is submitted as boundary conditions in FLUENT and boric acid mixing is analyzed to the core inlet.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A053. doi:10.1115/ICONE22-30417.

Ledinegg instability is one of the most important static instabilities for two phase flow system, especially in microchannel systems. In this paper, the force circulation two phase flow instability in vertical multi-channel system is performed by the best estimate system computer code RELAP5. The process and inherent reason of flow instability between multichannel system (FIBM) and flow excursion in forced circulation parallel channel system are analyzed. The effects of main operating parameters related to static onset of flow instability are investigated. Inlet subcooling, inlet restrictor, and saturation pressure are sensitive to the stability of parallel channel system.

Topics: Two-phase flow
Commentary by Dr. Valentin Fuster
2014;():V02AT09A054. doi:10.1115/ICONE22-30428.

Researches of steady-state thermal hydraulic performance of a large power PWR had been made using the COBRA III C/MIT-2 code. The basic thermal hydraulic parameters and characteristics of the core channels were studied. The results showed that the hottest channel and the maximum enthalpy rise hot channel were not boiling. The result will lay the foundation for further studies in the thermal-hydraulic design of a large advanced PWR.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A055. doi:10.1115/ICONE22-30437.

Effect of regulation valves (RV) installation in high pressure injection system (HPIS) pipelines on the formation of reactor pressure vessel (RPV) thermal stress conditions was analyzed. Modernization is implemented at South-Ukrainian nuclear power plant (SUNPP) Unit 1 within the framework of life extension, which finished by the end of 2013. The main goal of the modernization is to expand the HPIS functionality for small leak accident and protection against the cold overpressurization due to flow rate and primary pressure effectively regulation. The thermal hydraulic model for RELAP5/mod3.2 code with detailed downcomer (DC) model and changes in accordance with modernization was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. Also, an algorithm for the RVs was developed. Applying of cooling water flow rate regulation avoids excessive primary cooling and, consequently, helps to preserve the RPV integrity and to prevent reaching through crack formation, which can lead to a severe accident.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A056. doi:10.1115/ICONE22-30440.

The thermal-hydraulic characteristics of Nuclear Power Plant (NPP) during Steam Generator Tube Rupture (SGTR) accident are of great concern. In this paper, the thermal-hydraulic characteristics of CP300 during SGTR accident with no operator actionsduring the first 30 min are investigated using the best estimate RELAP5/MOD3.4.

Modeling and nodalization of CP300 was executed, including the vessel, pumps, pressurizer, steam generations and necessary auxiliary systems. Some main transient parameters were obtained, such as Reactor Coolant Pump (RCP) coolant temperature and pressure, steam generator flow rate and pressure. The calculation results gives the sequences of the NPP during the SGTR as described below. As the tube rupture occurs, the primary pressure drops and secondary pressure increases. When the primary pressure drops down to the set-point of the scram, the control rods drop down and the power of the NPP begins to decrease, causing the primary coolant temperature to decrease. The primary pressure continues to drop. When it drops down to 10.78MPa, the High pressure Safety Injection System is put into operation and the accident is mitigated, even without operator’s actions.

In conclusion, the calculated results indicate that the key thermal hydraulic parameters of CP300 during the SGTR accident are in acceptable ranges and the accident is effectively controlled.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A057. doi:10.1115/ICONE22-30450.

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A058. doi:10.1115/ICONE22-30452.

Different deaerators have different structures, and the intensities of deaerating are different. But the deaerator models without considering the structure couldn’t show these aspects. In order to ensure the accuracy of simulation results, concrete structure of the deaerator should be taken into consideration. This paper carried out a mathematical model of horizontal type deaerator with constant nozzles and trays in nuclear power plants. It was built based on the structure of the segmentation deaerators and the process of heat transferring. This paper calculated the vapors condensation rate, it was based on heat transfer coefficient of direct contact condensation, and the heat transfer area of water film when working conditions were corresponding to the empirical formulas. While the working conditions were beyond the limits of empirical formulas, this paper would build models by conservation of energy, so the model could work under any working conditions. These models were installed in the simulation system of Qinshan Phase II., and were tested under variable power and accidental conditions. The testing results show that the models could fit for different steady working conditions. Compared with the outputs and the actual operating data, the error was small. Under the conditions of variable power and turbine tripping, changing of the parameters have the same trend curve with the actual operating data. Because of taking the concrete structure into consideration, these models can be performed better of the specific characteristics of the horizontal type deaerators with constant nozzles and trays in nuclear power plants, therefore the results are more accurate than the models without considering the structure.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A059. doi:10.1115/ICONE22-30455.

In order to obtain the hydraulic resistance characteristics of steam generator (SG) tube support plates (TSP), experimental as well as CFD studies have been carried out on both the single-phase and two-phase hydraulic resistances of various trefoil or quatrefoil orifice plates. Results show that with the increase of the Renylod number, the single-phase pressure drop coefficient decreases firstly and remains almost constant later. The single-phase pressure drop coefficient decreases with the increase of the chamfer radius of orifice or flow area. The two-phase pressure drops predicted by the empirical correlations are generally larger than the experimental results, while the pressure drops predicted by CFD software agree with the experimental data.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A060. doi:10.1115/ICONE22-30463.

With the development of science and technology, some important passive features have been used in nuclear reactors, one of which is passive containment cooling system (PCCS). In the system, steam condensation plays an important role in removing heat from the containment atmosphere during a postulated accident. It has been found that during most time of an accident, the gas regime in the containment will be under natural and mixed convection. Advanced pressurized water reactor (CAP1400), designed by State Nuclear Power Technology Corporation (SNPTC) in China, is one of Chinese national science and technology projects. Since the PCCS has been applied in CAP1400, the study of condensation with non-condensable gases under natural and mixed convection becomes necessary.

To have a deeper understanding on the phenomenon of condensation with non-condensable gases under natural and mixed convection, an experiment facility was set up by State Nuclear Power Technology Research & Development Centre (SNPTRD). The test section of the facility is a rectangular channel with one of the walls acting as a condensing plate. The effects of buoyancy force on steam condensation with non-condensable gases are investigated. Also, a CFD model is set up to simulate the process.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A061. doi:10.1115/ICONE22-30468.

Once-Through Steam Generator (OTSG) is widely used in nuclear reactor system due to its advantages of compactness. The heat transfer performance of DOTSG is studied in this paper. In order to minimize the DOTSG volume and reduce the pressure drop of coolant, the pitch of inner helical tube is optimized with Pontryagin Maximum Principle (PMP). The double-tube is divided to three regions according to the coolant phase in secondary side. With given heat transfer load, choosing a combination function of minimum tube length and minimum pressure drop constructed with linear weighted method as objective function, the pitch optimization proceeds from superheated region to boiling region, and then to sub-cooled region in sequence, using Maximum Principle and gradient method. Then the pitch and temperature distribution along the axis is obtained respectively. The results show that the optimal pitch keeps constant along the axial direction in sub-cooled region and superheated region, but varies in boiling region. In boiling region, compared with minimum tube length optimization, the optimal tube length is 6.4% longer while the pressure drop is 36.3% smaller; and compared with minimum pressure drop optimization, the optimal pressure drop is 29.1% larger while the optimal tube length is 4.6% smaller. With the optimal pitch, the temperature distribution is in agreement with the general physic rules, which proves the correctness and the feasibility of the Maximum Principle method used for the structural optimization of DOTSG in this paper.

Topics: Boilers , Optimization
Commentary by Dr. Valentin Fuster
2014;():V02AT09A062. doi:10.1115/ICONE22-30477.

Mixed convection heat transfer in heated tubes has been studied extensively in the past decades, which is widely used in various industrial fields such as cooling of a nuclear reactor core. The secondary flow, which is induced by buoyancy force, has been found in previous research to have profound influence on the heat transfer difference on circumferential position and occurrence of heat transfer deterioration in horizontal heated channel. Therefore, understanding the secondary flow velocity field has important implications to prevent heat transfer deterioration and ensure the safe operation of nuclear power plants. Numerical methods have been adopted in literature to analyze the complex interaction between secondary flow and heat transfer deterioration. However, to the knowledge of the author, experimental measurement of secondary flow in the radial cross-section of a horizontal tube does not exist.

In this paper, a novel measurement method, which combines the transparent heating and PIV (Particle Image Velocimetry) technology, has been adopted to experimentally investigate the secondary flow velocity field on the radial cross-section in a horizontal heated tube. The heat transfer deterioration mechanism is revealed through analysis of the distribution of secondary flow along circumference direction at low mass flow rates and high heat flux conditions. We found that the buoyancy force lead the hot fluid to rise along the tube wall from bottom to top. While the secondary flow is most intensive near the middle of the interface, the secondary velocities are high at the bottom of the cross-section, where the tube wall is well cooled by cold fluid descends from the central part of the cross-section. Near the top of the tube wall, the secondary velocities are very small and the thermal acceleration effect makes the fluid rise. As a result, the mixed convection of top and center part of cross-section is weak and heat is primarily transferred by conduction, which leads the occurrence of thermal stratification of fluid. Consequently, the thermal accumulation of fluid in the top leads to heat transfer deterioration. Moreover, thermal properties differences between Freon (FC-72) and water, especially the Prandtl number (Pr), make the occurrence of heat transfer deterioration much easier for FC-72 than water with same working conditions.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A063. doi:10.1115/ICONE22-30482.

The high accurate throat tap flow nozzle with four different diameter taps is developed and its discharge coefficients are measured in the Reynolds number range from 1.5×106 to 1.4×107 using the high Reynolds calibration facility of AIST,NMIJ. The discharge coefficient of a throat tap nozzle extrapolated according to ASME PTC 6 are confirmed to deviate 0.37% at Red=1.4×107 from the experimental results. The high accurate flow nozzle developed can reduce this extrapolation error of the discharge coefficient to high Reynolds numbers by using the equations of discharge coefficients, which is determined as a function of Reynolds number and tap diameter based on the experimental results of four different diameter taps. The error of extrapolated discharge coefficient using the derived equations is estimated to be less than 0.1% at Red=1.4×107. The present results show that the throat tap flow nozzle developed is expected to work as a high accurate flowmeter even under the extrapolation of the discharge coefficient toward high Reynolds numbers.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A064. doi:10.1115/ICONE22-30485.

In this paper the thermal-hydraulic characteristics of the primary loop of the Experimental Breeder Reactor (EBR-II), including the temperature and the flow characteristics of the core, the intermediate heat exchanger (IHX) and the experiment subassembly XX09 and XX10, were analyzed with the transient thermal-hydraulic code THACS. The THACS code contains the core, the pumps, IHX, the sodium pool and some other modules, and each module could operate separately. All of the primary–loop components are simulated one-dimensional, and in the core calculation the incompressible model for the single phase. The multiple-channel model is applied to simulate the core subassemblies, including the average, hot, XX09, XX10, the reflector and the blanket channels. The neutron physics is calculated with the point reactor kinetics, and the reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of core are considered. Two tests, namely the SHRT-17 and SHRT-45R tests, are simulated to validate our tools and models. The THACS simulation results show that the EBR-II type sodium cooled fast reactor could shut down automatically relying on inherent negative feedbacks in the two tests.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A065. doi:10.1115/ICONE22-30490.

The advanced pressurized water reactor (APWR) uses a passive safety system relying on heat removal by condensation to maintain the containment within the design limits of pressure and temperature. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. In this paper, the test pressurized vessel in the experimental test on steam condensation on the cold surface for CAP1400 is designed, and the structure pressure is calculated.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A066. doi:10.1115/ICONE22-30493.

A new model which describes the dynamics of a vertically stratified flow correctly within the limits of a single-pressure two-fluid model has been developed. The model is based on the modification of finite-differences of convective terms and pressure gradients taking into account a distinct interface.

We propose to use the vapor quality as a criterion for the onset of dryout. The choice of the criterion is based on the analysis of experimental and theoretical studies. To determine the boundary vapor quality we used the correlation xcr = 1.26·G0.2, which was found from experimental data fit.

A review of articles has shown that for today it is impossible to predict correct superheat value. Therefore the superheat value was determined as a parameter of the model from the experimental data of a particular simulated experiment. Thus a boiling up regime was selected. The model described in this paper allows us to calculate the boiling up of sodium under the superheat conditions as well as problems of the evolution of the vapor volume.

The verification of the models was done by using the SOCRAT-BN code [1]. SOCRAT-BN is a coupled code which consists of modules for calculation of damage and melting of a reactor’s core, thermohydraulic processes and neutron physics.

The models of vertical stratification, dryout and slug boiling of superheated sodium are described in details in this paper. Also we present the results of verification for the models within analytic tests and experimental data.

Commentary by Dr. Valentin Fuster
2014;():V02AT09A067. doi:10.1115/ICONE22-30522.

The reactor core is a complex system involving the reactor physics, thermal hydraulics and many other aspects. That means the distribution of the core power largely determines the profile of the thermal parameters, meanwhile the local thermal-hydraulics condition will in turn affect the neutronics calculation by moderator temperature effect and Doppler effects. Issues coupling the thermal-hydraulics with neutronics of nuclear plants still challenge the design, safety and the operation of LWR few years ago. Fortunately, the recent availability of powerful computer and computational techniques has enlarged the capabilities of making more realistic simulations of complex phenomena in NPPs.

The current study deals with the development of an integrated thermal-hydraulics/neutronics model for Qinshan phase II NPP project reactor for the analysis of specific plant transients in which the neutronic response of the core is important, application of RELAP5-HD making use of the Helios code to derive the macroscopic cross-sections. Based on the coupled model, the steady state calculation and the transient simulation, involving the abnormal operation mode with asymmetrical coolant flux and temperature on the inlet of reactor, have been performed. The results show that the values obtained from coupled code RELAP5-HD calculation are in good agreement with the available experimental data, and the calculated accident parameters curves can predict all major trends of the transient. Steady state and transient condition calculation results are in accordance with the theoretical analysis from the aspect of coupled thermal-hydraulics/neutronics, this demonstrated a successful best estimate coupled RELAP5-HD model of Qinshan phase II NPP reactor has been developed, and the established model will provide a good foundation for the further analysis of the primary loop. It also can be concluded that the more accurate CFD method coupling three dimensional neutron kinetics code based on neutron diffusion method are necessary for steady-state calculation and analysis of transient/accident conditions when asymmetrical processes take place in the core. It is worth mentioning that RELAP5-HD code has already programmed the human-machine interface and the interface for coupling with other code, hence RELAP5-HD code has a broad application prospect in PWRs safety analysis.

Commentary by Dr. Valentin Fuster

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