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ASME Conference Presenter Attendance Policy and Archival Proceedings

2014;():V005T00A001. doi:10.1115/ICONE22-NS5.
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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to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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

Innovative Nuclear Power Plant Design and New Technology Application

2014;():V005T16A001. doi:10.1115/ICONE22-30387.

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.

Commentary by Dr. Valentin Fuster
2014;():V005T16A002. doi:10.1115/ICONE22-30438.

The innovative concept of pressurized water lead-bismuth-cooled fast reactor (PLFR) has been proposed and studied based on the previous LFR concept: PBWFR. Primary pumps and steam generators that contact lead-bismuth coolant are eliminated. A feedwater is directly injected into the primary coolant of hot lead-bismuth eutectic (LBE) at the outlet of the reactor core under the pressure of 14 MPa as PWR. The specifications of PLFR system are discussed and presented from thermal-hydraulic point of view.

Commentary by Dr. Valentin Fuster
2014;():V005T16A003. doi:10.1115/ICONE22-30487.

Now in Russia a new NPP “Brest – OD - 300” is developing which is to become a head in a series of fast reactors cooled by liquid lead or lead-bismuth alloy. Russian abbreviation OD is translated as Pilot-Demonstrational. The ideas are developed to use in this power plant deaeratorless thermal circuit. Such the schemes are widely and successfully used for conventional power plants in Russia.

Deaeratorless thermal schemes are based on the use of direct contact low-pressure reheaters. These reheaters have deaeration ability. Such schemes improve power plants efficiency by 1.5 %. Some other deaerator functions are distributed among other elements of plant. In particular, the presence of an external, independent supply of water, which is available at nuclear power plants and much higher than the supply of water in the deaerator.

Danger of return flow exists for direct-contact low pressure reheaters. But its design allows to eliminate completely the possibility of dangerous return flow. Compliance with safety was verified by calculation and in the operated power plants.

Commentary by Dr. Valentin Fuster
2014;():V005T16A004. doi:10.1115/ICONE22-30526.

PRHR system for low temperature and low pressure pool-type LWR, AHR400 is designed by two-phase closed thermosyphon and experimental validation is conducted. AHR400 is dedicated only to heat generation used in seawater desalination and operation temperature and pressure. LBLOCA is not considered as DBA due to no pipeline in primary system. There are LOHS and SBO for DBAs and PRHR system reduces damage during DBAs. Design of the PRHR system follows Direct Reactor Auxiliary Cooling System (DRACS) type. Two-phase closed thermosyphon, which uses phase change of working fluid, is applied to the PRHR system and the heat transfer in thermosyphon are analyzed by thermal resistance calculation model. Experimental thermosyphon that has similar thermal condition with the thermosyphon in designed PRHR system was explored for validation. The results show that the evaporation model overestimates heat transfer rate on the evaporator region.

Commentary by Dr. Valentin Fuster
2014;():V005T16A005. doi:10.1115/ICONE22-30615.

The Fluoride-salt-cooled High-temperature Reactor (FHR) is an advanced reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Chinese Academy of Sciences (CAS)’s Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal hydraulic transient analyses of an FHTR using SINAP’s pebble core design as a reference case. A point kinetic model is calculated by developing a microcomputer code coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating basic transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that the SINAP’s pebble core design is an inherently safe reactor design.

Commentary by Dr. Valentin Fuster
2014;():V005T16A006. doi:10.1115/ICONE22-30630.

The development of a small-sized nuclear heat-only plant with maximized safety features dedicated to seawater thermal desalination was proposed to address both a serious water crisis and nuclear safety issues, which continue to be perennial problems. In this study, the feasibility of a dedicated nuclear heat-only desalination system for a target country was evaluated in comparison with a target nuclear thermal desalination system. First, the target country was selected, and its current energy and desalination status was investigated. The suitable nuclear desalination options for the target country were then selected. Finally, using corresponding analysis tools, performance and economic analyses were conducted for a dedicated nuclear heat-only desalination system and the target nuclear thermal desalination system. The results of the analyses indicate that operating the small-sized nuclear heat-only plant at low pressures coupled with a seawater thermal desalination plant will considerably improve both the safety and economy without a significant loss in desalination performance. In conclusion, the proposed dedicated nuclear heat-only desalination system is expected to have high potential for solving both problems.

Commentary by Dr. Valentin Fuster
2014;():V005T16A007. doi:10.1115/ICONE22-30716.

As one of the Generation-IV reactor concepts, lead-alloy-cooled advanced nuclear energy systems (LACANES) have been studied worldwide in order to utilize the advantages of good heat transfer properties, neutron transparency and chemical inertness with air and water. Since the Fukushima accident, the passive safety aspect of the LACANES is increasingly emphasized due to outstanding natural circulation capability. To investigate the thermal-hydraulic capability of LBE, an international cooperation has been performed under OECD/NEA program, under the guidance of the Nuclear Science Committee by a task force named as Lead Alloy Cooled Advanced Nuclear Energy Systems (LACANES) since 2007. This international collaboration had dealt with computational benchmarking of isothermal LBE forced convection tests in the phase I, and the working group published a guideline for using one-dimensional system codes to simulate LBE forced circulation test results from HELIOS loop. The phase II was started after that, to give an additional guideline in the case of natural circulation. NACIE, one of benchmarking targets for the phase II which is a rectangular-shape loop located at ENEA-Brasimone Research Centre, Italy. NACIE test results were benchmarked by each participant using their one-dimensional thermal-hydraulic codes, and they are to follow the guideline from the LACANES phase I for regions where hydraulic loss occurs. Due to the selection of hydraulic loss coefficient relations by users, the cross-comparison results of international participants showed some discrepancies and the estimated mass flow rates had 13% of maximum error. Also, the future R&D areas are identified.

Commentary by Dr. Valentin Fuster
2014;():V005T16A008. doi:10.1115/ICONE22-30802.

Ceramic heat pipes and heat pipe based heat exchangers are tailored for automatically heat removal and heat distribution in thermally, chemically and abrasive high stressed systems. The manufacture of silicon carbide heat pipes was carried out. These were filled with sodium or zinc and sealed by laser brazing using metallic and glassy solder materials. High-temperature performance tests revealed a stable operating regime for both ceramic heat pipes with sodium and zinc as working fluid, respectively. Specifically the heat transferred by a zinc filled heat pipe of 22 mm in diameter and 750 mm in length accounted for 600 W at a temperature difference of 400 K. Notably the internal heat transfer capacity of the working fluid was even higher however, the total heat transfer was limited by the external active heat transfer area of the heat pipe. In order to evaluate the long-term stability of the heat pipes, particularly with respect to the joining seam, manufactured heat pipes are currently being tested in long-term annealing experiments at a temperature of 1000 °C under a variety of corrosive atmospheres.

Commentary by Dr. Valentin Fuster
2014;():V005T16A009. doi:10.1115/ICONE22-30903.

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.

Commentary by Dr. Valentin Fuster
2014;():V005T16A010. doi:10.1115/ICONE22-30995.

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Commentary by Dr. Valentin Fuster
2014;():V005T16A011. doi:10.1115/ICONE22-31280.

Nuclear energy has become the main energy source in Korea, but the safety issues are being debated since the Fukusima accident. In order to guarantee the safety and reliability of nuclear power plants, the uncertainty of human errors are being minimized by utilizing innovative technologies for inspection and maintenance. KAERI has developed robotic systems to upgrade and ensure the safety of nuclear facilities, to detect unusual conditions of facilities through remote inspection and to prevent human workers from high dose radiation with efficient plant maintenance.

Commentary by Dr. Valentin Fuster

Student Paper Competition

2014;():V005T17A001. doi:10.1115/ICONE22-30004.

This study has launched a concept to measure real time two-dimensional temperature distribution non-invasively by a combination of electrical capacitance tomography (ECT) technique and Debye equation. The concept has two steps which are the relative permittivity calculation from the measured capacitance among the many electrodes by ECT technique, and the temperature distribution calculation from the relative permittivity distribution by Debye equation. ECT sensor with 8 or 12-electrode is designed to measure and visualize the cross sectional temperature distribution in heating water as a basic experiment and melting polycarbonate pellets as a main experiment. Consequently, it is found that the water capacitance is changed by 1.14×10−6F as every 1.0 degree Celsius water temperature change. Moreover, the images of the temperature distribution from the relative permittivity distribution are reconstructed at every time step during the polycarbonate melting process. The non-invasive temperature values by a combination of ECT technique and Debye equation were compared with the invasive temperature values by the thermocouples. The non-invasive values have a good agreement with the invasive values by approximate 5%.

Commentary by Dr. Valentin Fuster
2014;():V005T17A002. doi:10.1115/ICONE22-30007.

The purpose of this study is to investigate a control method of natural circulation flow of air by injection of helium gas. A depressurization accident by the primary pipe rupture is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the double coaxial duct connecting between a reactor core and an intermediate heat exchanger (IHX) module breaks, air is expected to enter the reactor pressure vessel from the breach and oxidize in-core graphite structures. Then, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In such condition, injection of helium gas into the channel by a passive method can prevent occurrence of the natural circulation flow of air in the reactor pressure vessel. Therefore, it is thought that oxidation of in-core graphite structures by air ingress can be prevented by establishing this method.

The experiment has been carried out regarding the natural circulation flow using a circular tube consisting of a reverse U-shaped type. The vertical channel consists of one side heated tube and the other side cooled tube. The experimental procedure is as follows. Firstly, the apparatus is filled with air and one vertical tube is heated. Then, natural circulation of air will be produced in the channel. After the steady state is established, a small amount of helium gas is injected from the top of the channel. The velocity, mole fraction, temperature of gas, and temperature of the tube wall are measured during the experiment. The results were obtained as follows. When the temperature difference between the both vertical tubes was kept at about 60K, the velocity of the natural circulation flow of air was measured about 0.17m/s. During a steady state, a small amount of helium gas was injected into the channel. When the volume of injected helium gas is about 5.7% of the total volume of the channel, the velocity of the natural circulation flow of air became around zero. After 810 seconds elapsed, the natural circulation flow suddenly produced again. The natural circulation flow of air can be controlled by injecting of helium gas.

Commentary by Dr. Valentin Fuster
2014;():V005T17A003. doi:10.1115/ICONE22-30014.

Earthquake is one of the most serious phenomena for safety of a nuclear power plant. Therefore, nuclear reactors were contracted considering structural safety for a big earthquake. In a nuclear reactor, the gas-liquid two-phase flow is the one of primary factor of the property and bubbly or plug flow behavior is important issue to evaluate of safety. However, the influence of an earthquake vibration on the gas-liquid two-phase flow inside the nuclear power plant is not understood enough. For example, the bubbly flow behavior under the flow rate fluctuation caused by the earthquake acceleration is not clear. It is necessary to clear the two-phase flow behavior under the earthquake conditions. To develop the prediction technology of two-phase flow dynamics under the earthquake acceleration, the detailed two-phase flow simulation code with an advanced interface tracking method, TPFIT was expanded to the two-phase flow simulation under earthquake accelerating conditions. In the present study, the objective is to clarify the behavior of the gas-liquid two-phase flow under the earthquake conditions. Especially, the bubble behavior in the two-phase flow, a diameter, shape and velocity of bubbles which are expected to be influenced by the oscillation of the earthquake is investigated. In this experiment, the flow was bubbly flow and/or plug flow in a horizontal circular pipe. The working fluids were water and nitrogen gas. The nitrogen gas from gas cylinder was injected into the water through a nozzle and bubbly flow was generated at a mixer. The water was driven by a pump and the flow rate fluctuation was given by a reciprocating piston attached to the main flow loop. Main frequency of earthquakes is generally between 0.5Hz and 10Hz. Thus the frequency of the flow rate fluctuation in the experiment also was taken between 0.5Hz and 10Hz. The behavior of horizontal gas-liquid two-phase flow under the flow rate fluctuation was investigated by image processing using a high-speed video camera and PIV at test section. The pressure sensors were installed at the inlet of the mixer and the outlet of the test section. As the result, the bubble behavior mechanism under the flow rate fluctuation was obtained. In addition, the acceleration of a bubble and the pressure gradient in the pipe was synchronized under all frequency conditions. The prediction results by TPFIT were compared with the experimental results. They show good agreement on the flow field around a bubble and the bubble behavior.

Commentary by Dr. Valentin Fuster
2014;():V005T17A004. doi:10.1115/ICONE22-30016.

In a nuclear power plant, one of the important issues is an evaluation of the safety of the reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs.

However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there is little knowledge how coolant in a core response when large earthquake acceleration is added. Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction and/or the power of core are fluctuated with the oscillation by the experiments and numerical analysis. However the detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow behavior. In order to investigate the influence of vibration of construction on bubbly flow behavior, we visualized bubbly flow in pipeline on which sine wave was applied. In a test section, bubbly flow was produced by injecting gas into liquid flow through a horizontal circular pipe. In order to vibrate the test section, an oscillating table was used. The frequency and acceleration of vibration added from the oscillating table was from 1.0 Hz to 10 Hz and . 0.4 G (1 G=9.8 m/s2) at each frequency. The test section and a high speed video camera were fixed on the oscillating table. Thus the relative velocity between the camera and the test section was ignored. PIV measurement was also conducted to investigate interaction between bubble motion and surround in flow structure. Liquid pressure was also measured at upstream and downstream of the test section. The effects of oscillation on bubbly flow were quantitatively evaluated by these pressure measurements and the velocity field. In the results, it was observed that the difference of bubble motion by changing oscillation frequency. Moreover it was suggested that the bubble deformation is correlated with the fluctuation of liquid velocity field around the bubble and the pressure gradient in the flow area. In addition, these experimental results were compared with numerical simulation by a detailed two-phase flow simulation code with an advanced interface tracking method, TPFIT. Numerical simulation was qualitatively agreed with experimental results.

Commentary by Dr. Valentin Fuster
2014;():V005T17A005. doi:10.1115/ICONE22-30028.

Mitigative measures against a Core Disruptive Accident (CDA) are important from the viewpoints of safety of a Fast Breeder Reactor (FBR). If a CDA occurs, Post Accident Heat Removal (PAHR) must be surely achieved. In the PAHR, molten materials are likely to be injected into the coolant like a jet and they must satisfy two requests simultaneously: fast ejection and stable cooling after quenched. In order to estimate the quench behavior of the molten jet, it is important to understand how the jet breaks up.

The objective of this study is to clarify that the influence of hydrodynamic interaction between a jet and the surrounding fluid on jet breakup. Previous works have clarified that one cause of the jet breakup is provoked by fragmentation at the side of a jet. However, there are few detailed results describing the correlation between jet breakup and hydrodynamic interaction at the leading-edge region of a jet. Additionally, air entrainment with a jet is always observed in our past experiments using simulants, but its influence has not been discussed yet.

In this study, jet injection experiments in liquid-liquid system were conducted for investigating the interaction a jet and an ambient fluid, and the effect of air entrainment on jet breakup behavior. Both simulant core materials and coolants were transparent liquids for visualization. The stored simulant core material was injected into a tank filled with the simulant coolant. In order to realize the condition without air entrainment, the air remaining within the nozzle was removed using a syringe. The jet breakup behavior was observed with a high speed video camera. A normal backlight system and a Laser Induced Fluorescence (LIF) system were employed for visualization. The inner velocity distribution of a jet was measured by Particle Image Velocimetry (PIV).

As a result, in the experiments without air entrainment the jet breakup lengths were described by Epstein’s equation. In addition, a pair of vortices was observed at the leading-edge region. The vortices were generated at the leading edge and the leading edge rolled up by the vortices returned toward a jet core. Thus, it was very likely that the vortices at the leading edge region promoted jet breakup.

Commentary by Dr. Valentin Fuster
2014;():V005T17A006. doi:10.1115/ICONE22-30035.

On March 11, 2011, severe accident occurred at Fukushima Daiichi Nuclear Power Plant, and Units 1 to 3 of the plant have led to core melt. That is to say, melted fuel rods and core internals fell to the bottom of the Reactor Pressure Vessel (RPV). It is also believed that molten core has leaked into the reactor containment vessel. In order to plan for a safe molten core removal from the reactor, it is important to estimate the conditions of molten core by conducting analysis. Particular importance of the analysis is to understand the mechanisms of molten core spreading-cooling processes. However, sufficient understanding of this process has not been obtained yet.

The main purpose of this study is to evaluate molten metal spreading-cooling phenomena and subsequently, estimate the conditions of the molten metal. In order to achieve the purpose, the Computational Fluid Dynamics (CFD) for thermal fluid analysis, STAR-CCM+ was utilized. In the simulation of the unsteady two-phase flow, the volume of fluid model was applied for the spreading and interfacial surface formation of molten metal with the surrounding air. The key parameter for the molten metal spreading is the temperature dependent viscosity of molten metal. To assess the validity of this model, the analysis of the VULCANO VE-U7, molten metal spreading experiment, has been compared with simulation results.

Topics: Cooling , Accidents
Commentary by Dr. Valentin Fuster
2014;():V005T17A007. doi:10.1115/ICONE22-30040.

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical-energy generation. The largest group of operating Nuclear Power Plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) have gross thermal efficiencies ranging from 30% and up to 36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8 MPa / 257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fuel – natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants.

A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., SuperCritical Water-cooled Reactors (SCWRs) have to be designed. This path of the thermal-efficiency increasing is considered as a conventional way through which coal-fired power plants gone more than 50 years ago.

Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel Pressure-Channel (PCh) SCWR.

Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulic code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the CFD Fluent code has been used for better understanding of specifics of heat transfer in supercritical water.

Future studies will be dedicated to materials and fuels testing in an in-pile supercritical-water loop and developing passive-safety systems.

Commentary by Dr. Valentin Fuster
2014;():V005T17A008. doi:10.1115/ICONE22-30054.

At the beginning of the French nuclear history, CEA controlled all the nuclear development including its safety. In 1950s, this situation was changed by the participation of many industrial companies, which means industrialization of nuclear technology. This change became clearer when they adopted PWR in 1970. And so they needed and established a system to make safety regulation by putting the multiple actors’ opinions together.

After the accident of Chernobyl, antinuclear public opinion has increased. And in 1997, The Greens obtained the post of the Minister of Ecology. These facts required more independent and transparent regulatory system, so in 1998, a report which proposed to establish a new regulatory organization was published. On the basis of this report, they founded ASN in 2006.

From this French history, we can say that as the relationship between nuclear technology and the society changed, the regulatory body also changed to meet the necessity.

Commentary by Dr. Valentin Fuster
2014;():V005T17A009. doi:10.1115/ICONE22-30082.

The Accelerator Driven System (ADS) is a kind of nuclear reactor which can burn minor actinide waste products from conventional reactors with inherent safety. Because of the characteristics of a sub-critical reaction process, the fission chain reaction is maintained by additional neutrons generated by protons in spallation target. In this paper, a helium-cooled solid target is designed for a 10MW helium-cooled prismatic-type experimental ADS. The spallation material tungsten is modeled into the honeycomb structure, and high pressure helium flow in honeycomb holes remove heat deposited by the proton beam in the target. Since the complexity of geometry, if analysis the target with CFD code the computing time is unacceptable. Thus, a simple mathematical method was developed to solve the heat transfer problem in honeycomb structure, which avoids fine grid and the turbulent simulation, and gets a solution quickly. In order to analyze the transient characteristics of spallation target cooling system, a equivalent RELAP5 model for target and a RELAP5 model for cooling system are established. Results obtained indicate that the peak temperature in the target is lower than the limiting value under operating state and blockage condition as well as three typical transients with protected.

Commentary by Dr. Valentin Fuster
2014;():V005T17A010. doi:10.1115/ICONE22-30086.

One of the important severe accident management measures in a Light Water Reactor (LWR) is water injection to the reactor core. Reflooding of the uncovered reactor core is essential to prevent total core degradation. The series of QUENCH tests have been conducted to acquire knowledge of the reflooding. A number of analyses on QUENCH tests have also been done by different computer codes for code validation and improvements.

In this study, the modeling of the QUENCH-06 experiment was performed with RELAP/SCDAPSIM computer code. The input deck was modified and the SCDAP model was improved in order to represent the experimental facility more precisely. The uncertainties regarding the electrical resistance distribution in the heater rod system and the thermal properties of the shroud insulator were assessed, respectively.

The main purpose of this study is to identify the models to be improved. The rather good agreement between the calculation results and the measurement was acquired than the past studies [1]. In addition, more accurate modeling of the electrical resistance and the thermal properties of shroud insulator was done and their importance was indicated. The temperature profile and oxide thickness still showed similar tendencies with the original case. Further improvements are required mainly in the heat transfer model and the oxidation model in the SCDAP code.

Commentary by Dr. Valentin Fuster
2014;():V005T17A011. doi:10.1115/ICONE22-30104.

This paper reports the critical heat flux (CHF) enhancement that was observed experimentally when a porous metal was placed in a small flow channel (hereafter, this channel is called a “porous microchannel”). In the porous microchannel, the CHF value increased almost linearly with increased values of the mass flux and the inlet subcooling. In consequence, higher cooling performance was achieved under high mass flux and high inlet subcooling conditions. It was also found that considerable fluctuation of the pressure loss frequently encountered in a small heated channel disappears in the porous microchannel. It was considered that the stabilization of the pressure loss can mainly be attributed to inhibition of the formation of large bubbles. The effects of the material and the pore size of the porous metal were also investigated. Silver and nickel were selected as the porous metal material and the pore size tested was 0.2 and 0.6 mm. In the present experiments, the CHF value was not influenced significantly by the material in spite of the distinct difference of the thermal conductivity between silver and nickel, whilst it was dependent noticeably on the pore size. It was hence suggested that the CHF enhancement observed in this work was mainly caused by the complex thermal-hydraulic field formed in the porous microchannel. Preliminary results of the flow visualization performed to reveal the mechanisms of the CHF enhancement in the porous microchannel was also reported.

Commentary by Dr. Valentin Fuster
2014;():V005T17A012. doi:10.1115/ICONE22-30109.

The Modular High Temperature Gas-cooled Reactor (MHTGR) could realize higher efficiency and lower costs by developing the multi-modular high temperature gas-cooled reactors combined with supercritical steam turbine unit. The coupling effects among different modules are crucial to the designs and operation analyses of the multi-modular reactors. By establishing the engineering simulator for multi-modular reactors, the coupling effects can be studied and optimized to advance the reactor designs, due to the advantages of real-time calculations and coupled calculations. As key energy transfer equipment, the steam generator is very important to the reactor operation, and focused in the modeling of the engineering simulator system for multi-modular reactors.

In this paper, the once-through steam generator consisted of helical coils was modeled and optimized in the vPower integrated simulation platform. From the detailed analyses of the distributions of temperatures, heat flux, and other parameters along the heat transfer tubes, it showed that the steam generator model well presented the supercritical water properties and heat transfer characteristics inside helical tubes. Also, the heat transfer correlations of the supercritical water inside helical tubes were investigated, discussed and also compared to test the uncertainty and influence to the whole steam generator model. And the results indicated that most heat transfer correlations showed similar results and had little effect on the primary side in the steady state operation condition. In future work, the model and heat transfer characteristics of the supercritical steam generator will be further tested in more transients and integrated into complete engineering simulator for multi-module reactors.

Topics: Boilers , Modeling
Commentary by Dr. Valentin Fuster
2014;():V005T17A013. doi:10.1115/ICONE22-30122.

The appropriate description of heat-transfer to coolants at supercritical state is limited by the current understanding. Thus, poses one of the main challenges in development of supercritical-fluids applications for the Generation–IV reactors. The objective of the paper is, therefore, to discuss the basis for comparison of relatively recent experimental data on supercritical carbon dioxide (CO2) obtained at facilities of the Korea Atomic Energy Research Institute (KAERI) and Chalk River Laboratories (CRL) of Atomic Energy of Canada Limited (AECL). Based on the available instrumental error, a thorough analysis of experimental errors in wall- and bulk-fluid temperatures, and heat transfer coefficient is conducted. It is shown that rarely published data on instrumental errors tend to underpredict significantly actual experimental errors. A revised heat-transfer correlation for the CRL data is presented. A preliminary heat-transfer correlation for joint CRL and KAERI datasets is developed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A014. doi:10.1115/ICONE22-30125.

Subcooled boiling flows are encountered in most nuclear reactor configurations. Wall heat flux partitioning models form an integral part of the subcooled boiling formulations in CFD codes. These models attempt to describe the flow of heat from the wall into the fluid by dividing it according to several mechanisms of heat transfer. This work presents a one-dimensional evaluation of the wall heat flux partitioning model of Kurul and Podowski, also referred to commonly as the RPI model, which is used in the state-of-the-art codes of today. This model was assessed against the measurements of Okawa et al. for a vertically upward subcooled boiling flow of water at near atmospheric pressure. Although the predictions showed good agreement with the measured wall temperatures, significant discrepancies were observed in the predictions of the constituent sub-models that comprised the overall model. Prospects for improvement are discussed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A015. doi:10.1115/ICONE22-30132.

Current generation water-cooled Nuclear Power Plants (NPPs) have significantly lower thermal efficiencies than their thermal counterparts; due, partially, to their lower turbine-inlet steam temperature. Nuclear steam superheat can be implemented in a generic pressure-channel nuclear reactor to increase the temperature of the steam at the inlet of the turbine, and thus increase the thermal efficiency of a NPP.

A heat flux is computed specifically for a stable SuperHeated Steam (SHS) and Pressurized Water (PW) 520 pressure-channel reactor core configuration, from which a unique temperature profile for each coolant (as a bulk fluid) is calculated. Using the coolant temperature profile of each coolant, the sheath temperature distribution is calculated, using Fourier’s law, and the fuel pellets’ axial and radial temperature profiles are determined using an analytical solution to the temperature distribution in a solid with uniform heat generation.

Properties of the coolant, sheath, and fuel were calculated based on the temperature (and pressure, in the case of coolant) along the heated length of a channel. The effects on the flow rates and the differences in the required channel powers, due to the addition of the SHS channels, were also considered. To ensure safe operating parameters, the maximum sheath and fuel centerline temperatures were shown to be much lower than the operating limits.

The implementation of steam superheat in a generic 1200-MWel pressure-channel nuclear reactor allows for an increase in the temperature of steam at the inlet of a turbine from ∼319°C to ∼550°C, and ultimately an increase in the thermal efficiency of the NPP by about 5–7%.

Topics: Pressure , Steam
Commentary by Dr. Valentin Fuster
2014;():V005T17A016. doi:10.1115/ICONE22-30134.

The objective of this paper is to modify the current existing CANFLEX® fuel-bundle design to examine its ability to withstand high-temperature conditions of a proposed generic reactor with nuclear steam reheat. One of this reactor’s characteristics is having Super-Heated Steam (SHS) channels in addition to Pressurized-Water (PW) channels in order to increase the thermal efficiency of the plant by about 7–12%. This increase may be attained by raising the outlet temperature of the SHS-channels coolant to about 550°C. Operating at the higher temperatures will definitely have an effect on the mechanical and neutronic properties of fuel-channel materials, specifically on fuel-sheath and pressure-tube materials.

This paper compares Inconel-600 and SS-304 in order to determine the most suitable material for SHS-channel’s sheath and pressure tube. This is achieved by comparing strength of materials by performing stress- and displacement-analysis simulation using NX8.5 software (NX8.5, 2009).

The analysis in this paper can also be applied to other Nuclear-Power Plants (NPPs) that require operating at higher temperatures such as Super-Critical Water-cooled Reactors (SCWRs).

Commentary by Dr. Valentin Fuster
2014;():V005T17A017. doi:10.1115/ICONE22-30135.

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet.

One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors.

Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

Topics: Pressure , Steam
Commentary by Dr. Valentin Fuster
2014;():V005T17A018. doi:10.1115/ICONE22-30136.

SuperCritical Water-cooled Reactors (SCWRs) are one of six Generation-IV nuclear-reactor concepts. They are expected to have high thermal efficiencies within the range of 45–50% owing to the reactor’s high pressures and outlet temperatures. Efforts have been made to study the supercritical phenomena both analytically and experimentally. However, codes that have been used to study the phenomena analytically have not been validated for supercritical water.

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid- and transport-properties packages, thus enabling a transition from subcritical to supercritical fluid states. This extension needs to be validated using experimental data.

In this work, the applicability of ATHLET code to predict supercritical-water behaviour in various heat-transfer conditions is assessed. Several well-known heat-transfer correlations for supercritical fluids are added to the code and applied for the first time in ATHLET simulations of experiments. A numerical model in ATHLET is created to represent an experimental test section and results for the heat transfer coefficient, bulk fluid temperature, and the tube inside-wall temperature are compared with the experimental data.

The results from the ATHLET simulations are promising in the Normal and Enhanced Heat-Transfer Regimes. However, important phenomena such as Deteriorated Heat Transfer are currently not accurately predicted. While ATHLET can be used to develop preliminary design solutions for SCWRs, a significant effort in analysis of experimental work is required to make further advancements in the use of ATHLET for SCW applications.

Topics: Water
Commentary by Dr. Valentin Fuster
2014;():V005T17A019. doi:10.1115/ICONE22-30137.

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. These high temperature, high pressure reactors will have much higher operating parameters compared to current Nuclear Power Plants (NPPs) (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. In support of developing these SCWRs, studies are currently being conducted for heat transfer at supercritical conditions.

Currently, there are no experimental datasets for heat transfer at supercritical conditions from power-reactor fuel bundles to a coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare-tube data can be used as a conservative approach. A number of empirical generalized correlations, based on experimentally obtained datasets, have been proposed to calculate Heat Transfer Coefficients (HTCs) in forced convective heat transfer for various fluids, including water at supercritical pressures. There have been a number of methods applied to correlate heat transfer data. The most conventional approach is to modify the classical Dittus-Boelter correlation for forced convection. The Bishop et al. correlation is an example of this type modification with an addition of an entrance-region term.

The Mokry et al. correlation (2009) was developed as a Dittus-Boelter-type correlation with thermophysical properties taken at a bulk-fluid temperature. The derived correlation has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in normal and improved heat-transfer regimes. This correlation has an uncertainty of about ±25% for HTC values and about ±15% for calculated wall temperature. However, this correlation does not take into account the entrance-region effect.

The objective of this paper is an investigation of the entrance-region effect to be incorporated into the proposed Mokry et al. correlation (2009) in an attempt to further improve its accuracy.

Topics: Heat transfer , Water
Commentary by Dr. Valentin Fuster
2014;():V005T17A020. doi:10.1115/ICONE22-30146.

Currently, there are six Generation IV reactor systems under development worldwide: 1) Very-High-Temperature Reactor (VHTR); 2) Sodium-cooled Fast Reactor (SFR); 3) SuperCritical Water-cooled Reactor (SCWR), 4) Gas-cooled Fast Reactor (GFR), 5) Lead-cooled Fast Reactor (LFR); and 6) Molten Salt Reactor (MSR). Of these six systems, Canada has decided to pursue the SCWR as its choice for a Generation IV reactor, with some research being conducted on the VHTR. One main objective of SCWRs is to increase the thermal efficiency of current nuclear power plants from the 30–35% range to approximately 45–50%. In order to accomplish this, SCWRs are being designed to operate well above the critical point of water at pressures of 25 MPa and reactor outlet temperatures up to 625°C. These operating conditions also make the SCWR, along with the VHTR and other Generation IV systems, suitable candidates to support thermochemical hydrogen cogeneration.

The design and operation of a facility capable of accurately and safely conducting experiments in supercritical water is a very expensive task. In order to facilitate our understanding of supercritical heat-transfer phenomena, modeling fluids such as carbon dioxide, refrigerants, ammonia and helium can be used to complement our knowledge of supercritical fluids. Some of these fluids, namely helium and carbon dioxide, have also been considered as potential working fluids in some special designs of reactors and power cycles.

The objective of this paper is to investigate the feasibility of using alternative working fluids such as helium and Refrigerant-134a (R-134a) by comparing the fluid and transport properties with those of water. Operating conditions of SCWRs are scaled into those of the modeling fluid, R-134a, in order to provide proper SCWR-equivalent conditions. The equivalent properties for helium, which is one possible coolant for the VHTR, are also discussed. The thermophysical properties for selected working fluids are obtained from NIST REFPROP software. The results indicate that the thermophysical properties of the fluids undergo significant changes within the critical and pseudocritical regions similar to that of supercritical water. A sensitivity analysis for the effect of temperature on selected thermophysical properties at various supercritical pressures was performed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A021. doi:10.1115/ICONE22-30151.

It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living. In general, electrical energy can be generated by: 1) non-renewable-energy sources such as coal, natural gas, oil, and nuclear; and 2) renewable-energy sources such as hydro, wind, solar, biomass, geothermal and marine. However, the main sources for electrical-energy generation are: 1) thermal - primary coal and secondary natural gas; 2) “large” hydro and 3) nuclear. The rest of the energy sources might have visible impact just in some countries.

Modern advanced thermal power plants have reached very high thermal efficiencies (55–62%). In spite of that they are still the largest emitters of carbon dioxide into atmosphere. Due to that, reliable non-fossil-fuel energy generation, such as nuclear power, becomes more and more attractive. However, current Nuclear Power Plants (NPPs) are way behind by thermal efficiency (30–42%) compared to that of advanced thermal power plants. Therefore, it is important to consider various ways to enhance thermal efficiency of NPPs.

The paper presents comparison of thermodynamic cycles and layouts of modern NPPs and discusses ways to improve their thermal efficiencies.

Commentary by Dr. Valentin Fuster
2014;():V005T17A022. doi:10.1115/ICONE22-30158.

Transient forced convection heat transfer due to exponentially increasing heat input to a heater is important as a database for safety assessment of the transient heat transfer process in a Very High Temperature Reactor (VHTR). The knowledge of heat transfer enhancement using a heater with twisted configuration is also important for the high performance design of intermediate heater exchanger (IHX) in VHTR system. In this study, forced convection transient heat transfer for helium gas at various periods of exponential increase of heat input to a short thin twisted plate with various helix angles was experimentally studied. A forced convection heat transfer experimental apparatus was used to measure the experimental data. The test heater was mounted horizontally along the center part of a circular test channel. Twisted plates were made of thin platinum plate with a thickness of 0.1 mm and width of 2 mm and 4 mm. The heat generation rates of the heater were controlled and measured by a heat input control system. The heat generation rate, , was raised with exponential function, = Q0exp(t/τ). Where, t is time, and τ is period of heat generation rate. The mean temperature of the test heater was measured by resistance thermometry. The heat flux was obtained by the energy conservation equation. The test heater surface temperature was calculated from heat conduction equation of the heater. The transient heat transfer experimental data were measured for the periods ranged from 80 ms to 17 s and at a gas temperature of 303 K under 500 kPa. The flow velocities ranged from 4 m/s to 10 m/s. In the experiments, the twisted plates with different width were tested. The surface temperature and heat flux are increasing exponentially with the time. It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period longer than about 1 s, and it becomes higher for the period shorter than about 1 s. The heat transfer coefficients for total length of the twisted plate were compared with the values of flat plate which has the same width and thickness with the twisted one. The local mean heat transfer coefficients have been tested as well. The heat transfer coefficients of twisted plate are about 10% for 2 mm-width one and15% for 4 mm-width one higher than those of flat plate with same width at the quasi-steady state. And also, the heat transfer coefficients for the first half pitch are 24% higher than that for the total length of the same twisted plate. Therefore, an enhancement in the heat transfer coefficient for the twisted plate was clarified.

Commentary by Dr. Valentin Fuster
2014;():V005T17A023. doi:10.1115/ICONE22-30168.

The two-phase natural circulation cooling performance of the APR1400 core catcher system is studied utilizing a drift flux flow model developed via scaling analysis and with an air-water experimental facility. Scaling analysis was carried out to identify key parameters, so that model facility could simulates two-phase natural circulation. In the experimental apparatus, instead of steam, air is injected into the top wall of the test channel to simulate bubble formation and void distribution due to boiling water in the core catcher channel. Measurement of void fraction critical to the heat transfer between the wall and coolant is carried out at certain key position using double-sensor conductivity probes. Results from the model provide expected natural circulation flow rate in the cooling channel of the core catcher system. The observed flow regimes and the data on void fraction are presented. For a given design of the down comer piping entrance condition bubble entrainment was observed that significantly reduced the natural circulation flow rate.

Topics: Cooling , Safety
Commentary by Dr. Valentin Fuster
2014;():V005T17A024. doi:10.1115/ICONE22-30174.

Betavoltaic cells can provide extended power up to 10 or more years in extreme temperature environments, −55°C to 150°C. However there is limited study on the loading of tritium which is beta source for these cells. The present study examines the loading of the tritium using surrogate hydrogen gas in various films through experiments and simulations. A detailed review of the betavoltaic cell characteristics is first discussed and key challenges in this technology are identified. For the experimental work, a testing facility is designed for loading hydrogen in metallic films such as titanium, palladium and scandium which are good for storage of hydrogen or tritium. The facility is unique as it enables precise measurement of hydrogen loading in the films using pressure difference. Preliminary tests of loading on scandium films were carried out and some results are presented. In order to optimize the film thickness simulations were carried out using MC-SET code for beta flux emission. The results of the simulations for titanium and palladium film are presented.

Commentary by Dr. Valentin Fuster
2014;():V005T17A025. doi:10.1115/ICONE22-30176.

The China EU cooperation project SCWR-FQT small experimental reactor is selected as the research object. The new SCWR-FQT calculation code is made. Thermal performance of four channels are analyzed. The results show that: During normal condition, the coolant temperature increases gradually and pressure reduces gradually. The highest temperature of the coolant does not exceed 450°C working limit, meets the design safety requirement. When transient of entrance temperature occurs, lateral two channels are affected significantly. The inside channel has maximum temperature stability because of the minimum affection. The increasing of the entrance temperature will shorten the time of the relative inner two channels to reach stable. When transient of entrance flow occurs, the inside channel has the maximum affection. The decreasing of the entrance flow will increase the time of the inside channel to reach stable. The entrance flow should not lower than 49.2% of the normal operating mode.

Commentary by Dr. Valentin Fuster
2014;():V005T17A026. doi:10.1115/ICONE22-30180.

In line with the cooperative project “Non-invasive Condition Monitoring of Nuclear Reactors for the Detection of Level Change and Deformation of the Core” between the Technical University Dresden and the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of Zittau/Goerlitz University of Applied Sciences, a measuring system for the core state diagnosis during a core melt accident in the reactor pressure vessel of a pressurized water reactor is going to be developed. The operational principle of this system is based on the non-invasive measurement of continuously changing gamma radiations (caused by the shifting of melted materials and fission products) outside of the reactor pressure vessel by means of several gamma radiation sensors. The sensors are arranged over the height of the core and the lower head. By using computer based and real-time capable methods for evaluation of the measured gamma radiations conclusions about the core state can subsequently be drawn. This paper includes a description of a concept as well as several methods for the core state diagnosis during a core melt accident. The main part is the analysis of the methods for the core state diagnosis based on different evaluation criteria.

Commentary by Dr. Valentin Fuster
2014;():V005T17A027. doi:10.1115/ICONE22-30193.

The development of selective adsorbents has become very important for the effective multi-nuclide decontamination. In this study, the selective adsorption properties of 26 nuclides for different types of zeolites (A, L, natural mordenite (NM), Ag-NM) were examined in the presence of boric acid.

The batch adsorption experiments were carried out using four kinds of test solutions containing boric acid and calcium hydroxide; (1)DW (distilled water) + H3BO4: 3,000 ppm + LiOH: 10 ppb, (2)DW + Ca(OH)2: 500 ppm + H3BO4: 3,000 ppm + LiOH: 10 ppb, (3)Seawater (30% diluted) + H3BO4: 3,000 ppm, (4)Seawater + H3BO4: 3,000ppm. The uptake (%) of Sr2+ for zeolite A (A-51J), Cs+ for natural mordenite (NM, 2460#, Ayashi, Sendai), and I for Ag-NM was determined under the following conditions; Concentration of Sr2+, Cs+ and I ions: 10 ppm, V/m = 100 cm3/g, 25°C, 24 h. The uptake (%) of Sr2+, Cs+ and I ions was estimated to be above 90%, while tended to decrease in the presence of seawater. Especially, the uptake (%) of I ions for Ag-NM markedly decreased in the presence of seawater.

As for the zeolites A and L, the uptake (%) of 26 elements was determined by using two kinds of test solutions; (1)DW (distilled water) + H3BO4: 3,000 ppm + LiOH: 10 ppb + 26 nuclides: 10 ppm, (2)Seawater (30% diluted) + H3BO4: 3,000 ppm + 26 nuclides: 10 ppm.

Zeolite A has relatively large uptake percentage for Sr, Co, Ni and Zn, and zeolite L has high adsorbability to lanthanoid group of Eu, Ce and Pr. The increase in pH led to the enhancement of uptake (%), while the hydrolysis of metal ions should be also considered. The multi-nuclides separation is thus expected by considering the difference in uptake properties of zeolite A, L and natural mordenite.

Commentary by Dr. Valentin Fuster
2014;():V005T17A028. doi:10.1115/ICONE22-30208.

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 with actinides and full isotopic composition has been performed.

The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight.

As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system.

Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

Commentary by Dr. Valentin Fuster
2014;():V005T17A029. doi:10.1115/ICONE22-30219.

The Fluoride-salt-cooled High temperature Reactor (FHR) is new reactor concept-about a decade old which is mainly on going in China and U.S. The preliminary thermal-hydraulic studies of the Fluoride salt cooled High temperature Test Reactor (FHTR) is necessary for the development of the FHR technology. In this paper, the thermal-hydraulics of FHTR (also called TMSR-SF) designed by Shanghai Instituted of Applied Physics (SINAP) is studied in different power modes. The temperature distributions of the coolant and the fuel pebble are obtained using a steady-state thermal-hydraulic analysis code for FHR. The comprehensive local flow and heat transfer are investigated by computational fluid dynamics (CFD) for the locations where may have the maximum pebble temperature based on the results from single channel analysis. The profiles of temperature, velocity, pressure and Nu of the coolant on the surface of the pebble as well as the temperature distribution of a fuel pebble are obtained and analyzed. Numerical results showed that the results of 3-D simulation are in reasonable agreement with that of single channel model and also illustrated safety operation of the preliminary designed TMSR-SF in different power mode.

Commentary by Dr. Valentin Fuster
2014;():V005T17A030. doi:10.1115/ICONE22-30263.

The process of heat transfer in a heavy liquid-metal coolant (HLMC) cross-flow around heat-transfer tubes is not yet thoroughly studied. Therefore, it is of great interest to carry out experimental studies for determining the heat transfer characteristics in a lead coolant cross-flow around tubes. It is also interesting to explore the velocity and temperature fields in a HLMC flow.

To achieve this goal, experts of the R.E. Alekseev Nizhny Novgorod State Technical University performed the work aimed at the experimental determination of the temperature and velocity fields in high-temperature lead coolant cross-flows around a tube bundle.

The experimental studies were carried out in a specially designed high-temperature liquid-metal facility. The experimental facility is a combination of two high-temperature liquid-metal setups, i.e., FT-2 with a lead coolant and FT-1 with a lead-bismuth coolant, united by an experimental site.

The experimental site is a model of the steam generator of the BREST reactor facility. The heat-transfer surface is an in-line tube bank of a diameter of 17×3.5 mm, which is made of 10H9NSMFB ferritic-martensitic steel. The temperature of the heat-transfer surface is measured with thermocouples of a diameter of 1 mm being installed in the walls of heat-transfer tubes. The velocity and temperature fields in a high-temperature HLMC flow are measured with special sensors installed in the flow cross section between the rows of heat-transfer tubes.

The characteristics of heat transfer and velocity fields in a lead coolant flow were studied in different directions of the coolant flow: the vertical (“top-down” and “bottom-up” [1]) and the horizontal ones. The studies were conducted under the following operating conditions: the temperature of lead was t=450–500°C, the thermodynamic activity of oxygen was a=10−5−100, and the lead flow through the experimental site was Q = 3–6 m3/h, which corresponds to coolant velocities of V = 0.4–0.8 m/s.

Comprehensive experimental studies of the characteristics of heat transfer in a lead coolant cross-flow around tubes have been carried out for the first time and the dependences NU = f(Pe) for a controlled and regulated content of the thermodynamically active oxygen impurity and sediments of impurities have been obtained.

The effect of the oxygen impurity content in the coolant and characteristics of protective oxide coatings on the temperature and velocity fields in a lead coolant flow is revealed. This is because the presence of oxygen in the coolant and oxide coatings on the surface, which restrict the liquid-metal flow, leads to a change in the characteristics of the wall-adjacent region.

The obtained experimental data on the distribution of the velocity and temperature fields in a HLMC flow permit studying the heat-transfer processes and, on this basis, creating program codes for engineering calculations of HLMC flows around heat-transfer surfaces.

Commentary by Dr. Valentin Fuster
2014;():V005T17A031. doi:10.1115/ICONE22-30269.

The specific feature of heavy liquid-metal coolants (HLMC) is the possibility to form solid-phase impurity particles, which requires a deep study of characteristics of the wall boundary layer enriched with impurity particles. This is necessary for a fundamental understanding of the processes occurring on contact surfaces and triboprocesses to validate the use of materials for developing fast reactors with these coolants. The paper deals with results of experimental studies of structures and characteristics of the wall boundary layer. The use of the thermal shock technique enabled us to experimentally determine the wall boundary structure typical of circuits with heavy liquid-metal coolants (lead or lead-bismuth alloy). It has been experimentally demonstrated that the wall boundary region is a multicomponent structure:

1 – steel;

2 – oxide coating;

3 – layer of loose deposits weakly adhering to oxide coating;

4 – gas phase (due to unwettability of oxide surface by coolant);

5 – impurity-rich diffusion layer of boundary turbulent layer;

6 – boundary turbulent layer;

7 – impurity particles in coolant flow close to wall boundary region, which do not adhere to layers 3 and 5.

It is found that long-term HLMC circulation in channels leads to an increase in the surface roughness of constructional materials due to the deposition of solid-phase impurities, which should be taken into account in tribological studies.

The experimental results have also shown that the microhardness of structures of the wall boundary layer is an order different from the microhardness of both steel and solidified lead ingot (HVsteel = 245, HVwall boundary layer = 60, HVlead = 6), which permits to assume that the wall boundary layer is a rheological fluid.

Also the chemical analysis of deposits on construction material surfaces is presented. The chemical analysis included X-Ray method and elemental analysis.

Commentary by Dr. Valentin Fuster
2014;():V005T17A032. doi:10.1115/ICONE22-30274.

Taking into account the expected increase in global energy demands and increasing climate change issues there is a pressing need to develop new environmentally sustainable energy systems. Nuclear energy will play a big part in being part of the energy mix since it offers a relatively clean, safe and reliable source of energy. However, opportunities for building new generation nuclear systems will depend on their economic and safety attractiveness as well as flexibility in design to adapt to different countries and situational needs. Keeping these objectives in mind, a framework for international cooperation was set forth in a charter of Generation IV International Forum (GIF, 2002). The main design goals for the Generation IV nuclear system concept were to improve economic gains, enhance safety, extend sustainability, and strengthen proliferation resistance.

To achieve high thermal efficiencies of up to 45–50%, the use of SuperCritical Fluids (SCFs) as working fluids in heat transfer cycles is proposed. An important step towards development of SCF applications in novel Generation IV Nuclear Power Plant (NPP) designs and in other industries; is to understand the thermal hydraulic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions.

Heat transfer under SC turbulent conditions is generally very complex and is extremely sensitive to the test geometry and operational flow parameters. Detailed sets of experiments have been conducted around the world in tubes and other geometries with SCFs to study the basics of heat transfer. By variation of operational parameters and test geometries, fundamental heat transfer data sets are collected that will help in our understanding of SC heat transfer phenomena.

Applications for SCFs are not limited to power industry; recent advancements have indicated the use of SCFs in a much wider range of applications due to its unique and attractive heat transfer characteristics. In this paper, proposed uses of SCFs are presented and basic concept of DHT within SCFs is analyzed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A033. doi:10.1115/ICONE22-30289.

The development of selective adsorbents for radioactive Sr ions is one of the most important subjects for the safety decontamination in Fukushima NPP-1[1]. In this study, the selective adsorption properties of Sr, characterization and stable solidification were clarified by using the novel adsorbent of potassium titanates (KT-1).

The adsorption properties of Sr2+ ions for original and calcined specimens were examined by batch method under the following conditions; V/m = 100 cm3/g, Mixed solution: 10,000 ppm Na+, 10 ppm Cs+, 10 ppm Ca2+, 1 ppm Mg2+ and 1 ppm Sr2+, 85Sr tracer: 5,000 cpm/cm3, centrifugation: 2,500 rpm, 25°C, shaking time: 1∼24 h, calcination temp.: 300∼900°C. Relatively large uptake percentage above 90% was obtained for the original and calcined specimens below 800°C, while the Sr uptake for calcined specimens above 900°C was lowered due to the thermal decomposition of K2Ti2O5·xH2O.

The Sr distribution in the column was examined by flowing the mixed solution through the columns packed with KT-1. The Sr distribution profiles were obtained by the measurement of γ-activity in the column at 5 mm intervals. In either case, no breakthrough of Sr was observed. The distribution profile tended to smooth with increasing flow rate; Sr adsorption band and flow rate have a linear relationship.

The leachability of Sr for the solid forms was further examined under the following leaching conditions; leachant: pure water and 1 M HCl; leachant temp.: 25°C and 90°C, leaching period: 4 weeks; calcining temp.: 500∼1,100°C. The leached percentage of Sr in pure water was less than the detection limit of ICP-AES, and that in 0.1 M HCl tended to markedly decrease with calcining temperature; the formation of SrTiO3 phase above 800°C was effective for the lowering of leachability.

The novel adsorbent of KT-1 is thus effective for the selective decontamination and stable solidification of Sr in Fukushima NPP-1.

Commentary by Dr. Valentin Fuster
2014;():V005T17A034. doi:10.1115/ICONE22-30307.

Taking into consideration the unresolved resonance self-shielding effect, probability table is one of the most important and natural methods used to simulate the neutron transport in unresolved resonance range in reactor physics. A module named PURC for probabilistic unresolved resonance calculation which is embedded in the RXSP Beta2.0 code, has been developed by REAL team in Department of Engineering Physics in Tsinghua University. It is the optimization in sort algorithm that make PURC module more efficient. After applying the OpenMP parallel algorithm into this module, it has improved computational efficiency by more than one order of magnitude comparing with the corresponding functional module named PURR in NJOY code. Meanwhile the computational accuracy of PURC module is validated and verified by series of microscopic cross section comparisons and macroscopic criticality benchmarks.

Commentary by Dr. Valentin Fuster
2014;():V005T17A035. doi:10.1115/ICONE22-30317.

For advanced passive PWR, reactor coolant system (RCS) depressurization through automatic depressurization system (ADS) is an important measurement to avoid high-pressure melt ejection and direct containment heating. It allows injection from passive core cooling system and the implement of in-vessel retention. However, it has negative impact that hydrogen in the RCS can be released to the containment together with coolant, which may lead to hydrogen burning or even explosion in the containment. Therefore, this paper analyzes the RCS depressurization strategy during severe accident, and evaluates its negative impact. Severe accident sequences induced by station black out (SBO) was selected and analyzed with integral severe accident analysis code as a typical high pressure core melt accident scenario. Different depressurization strategies with ADS system were discussed based on Severe Accident Management Guideline (SAMG.) ADS valves were manually opened at a core exit temperature of 923 K with 20min delay for operator reaction. Both depressurization effect and hydrogen risk were evaluated for different strategies. Hydrogen distribution was calculated, which was used to determine the combustion mode in different compartments. Result shows all three strategies analyzed in this paper can depressurize the RCS effectively. And opening the ADS stage 1–3 valves causes rapidly increase of the hydrogen concentration in the in-containment refueling water storage tank (IRWST) compartment and may lead to hydrogen denotation. However, hydrogen can be well dispersed in the loop compartment with intentional open of ADS stage 4 valves to RCS depressurization. Therefore, suggestions are proposed for SAMG: implement RCS depressurization strategy with stage 4 ADS instead of ADS stage 1–3.

Topics: Hydrogen , Risk
Commentary by Dr. Valentin Fuster
2014;():V005T17A036. doi:10.1115/ICONE22-30325.

State of the art neutron detectors lack capabilities required by the fields of homeland security, health physics, and even for direct in-core nuclear power monitoring. A new system being developed at Purdue’s Metastable Fluid and Advanced Research Laboratory in conjunction with S/A Labs, LLC provides capabilities the state of the art lacks, and simultaneously with beta (β) and gamma (γ) blindness, high (> 90% intrinsic) efficiency for neutron/alpha spectroscopy and directionality, simple detection mechanism, and lowered electronic component dependence. This system, the Tensioned Metastable Fluid Detector (TMFD) [3], provides these capabilities despite its vastly reduced cost and complexity compared with equivalent present day systems. Fluids may be placed at pressures lower than perfect vacuum (i.e. negative) [4, 5], resulting in tensioned metastable states. These states may be induced by tensioning fluids just as one would tension solids. The TMFD works by cavitation nucleation of bubbles resulting from energy deposited by charged ions or laser photon pileup heating of fluid molecules which are placed under sufficiently tensioned (negative) pressure states of metastability. The charged ions may be created from neutron scattering, or from energetic charged particles such as alphas, alpha recoils, fission fragments, etc. A methodology has been created to profile the pressures in these chambers by lasing, called Laser Induced Cavitation (LIC), for verification of a multiphysics simulation of the chambers. The methodology and simulation together have lead to large efficiency gains in the current Acoustically Tensioned Metastable Fluid Detector (ATMFD) system. This paper describes in detail the LIC methodology and provides background on the simulation it validates.

Commentary by Dr. Valentin Fuster
2014;():V005T17A037. doi:10.1115/ICONE22-30433.

Experimental investigations on boiling heat transfer coefficients of boiling flows in rectangular narrow channel under rolling motion condition are performed. The cross section of the testing rectangular narrow channel is 2×40 mm, and the mechanical rolling thermal-hydraulic experimental facility is used in the experimental research of boiling heat transfer characteristics. Deionized water is used as the working fluid. The results show that the amplitude of boiling heat transfer coefficients of rectangular narrow channel increases with increasing rolling amplitude and rolling period of the rolling platform, the time average boiling heat transfer coefficients of test section in rolling motion are equal to the coefficients of the test section at equilibrium position, and with the increase of rolling amplitude and rolling period the time average boiling heat transfer coefficient almost unchanged. The amplitude of boiling heat transfer coefficients increases with increasing heat flux and flow rate, while decreases with the increase of system pressure. The curve of boiling heat transfer coefficient fluctuations of rectangular narrow channel is close to sine or cosine curve when the rolling period less than 15 seconds.

Commentary by Dr. Valentin Fuster
2014;():V005T17A038. doi:10.1115/ICONE22-30457.

A Field Programmable Gate Array (FPGA) is a type of integrated circuit (IC), which is programmed after it is manufactured. FPGAs are referred to as a form of programmable hardware, as there is typically no software or operating system running on the FPGA itself. A significant amount of design work has been performed regarding the application of FPGAs in the nuclear field in recent years, with much of that work centered around safety related Instrumentation and Control (I&C) systems and safety systems. These new FPGA based systems are considered to be viable alternatives to replace many old I&C systems that are commonly used in Nuclear Power Plants (NPPs). Many of these older analog and digital systems are obsolete, and it has become increasingly difficult to maintain and repair them. FPGAs possess certain advantages over traditional analog circuits, PLCs and microprocessors, when considering nuclear I&C and safety system applications. This paper describes how FPGA technology has been used to construct a lab-scale implementation of a Post-Accident Monitoring System (PAMS), for a Westinghouse AP1000 Nuclear Power Plant, using a National Instruments “cRIO” chassis and I/O modules. This system will perform the major functions of the existing PAMS, including monitoring the vital values such as temperature, water level, pressure, flow rate, radiation levels and neutron flux in the event of a serious reactor accident. These values are required in standards such as United States Nuclear Regulatory Commission (NRC), Canadian Nuclear Safety Commission (CNSC), International Electrotechnical Commission (IEC), and Institute of Electrical and Electronics Engineers (IEEE). All of the input signals are read and processed using the FPGA, which includes alarms if the values go beyond the specified range, or if the values change rapidly. The values were then output to the computer through the FPGA interface to provide information to the operator, as well as being sent through analog and digital output modules for further processing. The system was tested using both simulated and real inputs from sensors. Furthermore, the reliability of the new system has also been analyzed, using the Dynamic Flowgraph Methodology (DFM). DFM has been successfully applied in both the nuclear and aerospace fields, and has been described as one of the best methodologies for modeling software/hardware interactions, by the scientific literature as well as in NRC reports. DFM was applied to fine-tune the design parameters by determining the potential causes of faults in the design, as well as to highlight the effectiveness of DFM in nuclear and in FPGA applications.

Commentary by Dr. Valentin Fuster
2014;():V005T17A039. doi:10.1115/ICONE22-30495.

In the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), absorber sphere shutdown system is the second shutdown system. Absorber spheres are transported back to storage vessel by pneumatic conveying when reactor needs to be started. The restart reliability of absorber sphere pneumatic conveying is quite important. Experimental system for absorber sphere pneumatic conveying has been built to investigate the gas-solid flow characteristics of pneumatic conveying. Restart experiments were conducted to study the restart performance of pneumatic conveying. Ambient air was used as source gas and absorber sphere was replaced by glass sphere. The inlet steady gas velocity of the sphere feeder was in the range of 17∼27 m/s. Pneumatic conveying could be restarted in all the experiments, which showed the good reliability of the pneumatic conveying process. The maximum relative deviation of the spheres pile heights in the riser was 4.1%, which showed a good repeatability. Pressure drop of feeder during steady stage was in the range of 4∼8 kPa. Furthermore, the restart experiments were also conducted in the full scale experimental system with helium as the source gas and the pressure of 2 MPa, the results also showed the good reliability of pneumatic conveying process.

Commentary by Dr. Valentin Fuster
2014;():V005T17A040. doi:10.1115/ICONE22-30507.

One of the most crucial issues in development of the lead-bismuth cooled fast reactors (LFR), and the accelerator-driven system (ADS) is the compatibility of structural steels with liquid lead-bismuth eutectic (LBE) at high temperature range from 500°C to 650°C. In the primary circuit of LBE, metallic impurities such as Fe, Cr and Ni dissolve from structural steels into LBE at high temperature region, are transported to low temperature region, and precipitate as solid metals on channel wall. In this circulation, the data bases of diffusion coefficients of the metallic impurities are required for the estimate of mass transfer rates. In order to investigate the characteristics of diffusion of metallic impurities, the measurement of diffusion coefficients of Ni and Fe in liquid LBE was carried out using a capillary method. The concentrations of the diffusing species in LBE were determined by means of the inductively coupled plasma mass spectrometry (ICP-MS). From the measured axial distribution of Ni and Fe concentration in the LBE, the diffusion coefficient of Ni and Fe were obtained. The results are expressed by Display Formula

DNi=1.7×10-3exp-3.63×10-4/RT500°CT650°Ccm2/sDFe=3.5×10-3exp-4.15×10-4/RT550°CT650°Ccm2/s
It was found that the value of diffusion coefficient of Ni is similar to the diffusion coefficient of Fe in liquid LBE.

Commentary by Dr. Valentin Fuster
2014;():V005T17A041. doi:10.1115/ICONE22-30515.

The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents a study on the consequences of loss of heat removal accident in the spent fuel pool of a typical pressurized water reactor using the Modular Accident Analysis Program (MAAP5) code. Analysis of uncompensated loss of water due to the loss of heat removal with initial pool water level of 12.2 m (designated as a reference case) has been performed. The analyses cover a broad spectrum of severe accident in the spent fuel pool. Those consequences such as overheating of uncovered fuel assemblies, oxidation of zirconium and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission product are also analyzed in this paper. Furthermore, as important mitigation measures, the effects of makeup water in SFP on the accident progressions have also been investigated based on the events of spent fuels uncovery. The results showed that spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate higher than evaporation rate defined in the reference case. Although spent fuel assemblies partly exposed due to a mass flow rate of makeup water smaller than the average evaporation rate, continuous steam cooling and radiation heat transfer might maintain the spent fuels coolability as the actual evaporation was balanced by the makeup in a period of time of the order of several days. However, larger makeup rate should be guaranteed to ensure long-term safety of SFP.

Commentary by Dr. Valentin Fuster
2014;():V005T17A042. doi:10.1115/ICONE22-30558.

Since boiling heat transfer has a high heat transfer coefficient, it has been used as a cooling technique for high-temperature bodies and it has been investigated for more than 70 years [1]. However, it has not yet been fully clarified: because the boiling phenomena are affected by many factors, such as the coalescence of bubbles, the fluid convection, the heat conduction and the physics on the contact of the gas, the liquid and the solid phase, boiling phenomena are considerably complicated. The present paper investigated an effect of pressure on boiling heat transfer mechanism by using the MEMS technology. And, boiling heat transfer enhancement in water under low pressure and low boiling temperature was examined experimentally. Steady state pool boiling experiments were conducted by using a copper thin-film and a silicon wafer for the test heater and pure water at atmospheric condition for the test liquid. The system pressure was 0.010, 0.10 and 0.15 MPa, respectively. The heaters were made of a printed circuit board and a commercial silicon wafer. The width was 7.5 mm, the length was 10 mm. The test sections were arranged for horizontal position facing upward. The test heaters had an artificial cylindrical-cavity of 0.010 or 0.040 mm in diameter; the cavities were fabricated by using the MEMS technology, i.e., wet etching technology and deep RIE. The test heaters were heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The present experimental results showed the boiling bubble grew up to about 20 mm in diameter then the bubble released without coalesce of bubble under low pressure condition. Thus, the bubble coalesce was slight. From the experimental results, the gradient of boiling curve by using the copper thin-film was about 3: the heat transfer characteristic was dominant to nucleate boiling. On the other hand, the gradient of boiling curve by using the Silicon wafer (Non-cavity) was about unity: the heat transfer was dominant to heat convection of single-phase flow. According to the present observation of the boiling bubbles, the boiling heat transfer was dominant to latent heat: the ratio of the phase change and the convection was about 90 % and 10 %, respectively. The heat transfer ratio of the convection increased as the system pressure increased.

Commentary by Dr. Valentin Fuster
2014;():V005T17A043. doi:10.1115/ICONE22-30568.

Experiments of counter-current two-phase flow of upward steam flow and condensing downward film flow in a pipe were performed. The experiments were intended to examine water accumulation in steam generator U-tubes during intermediate and small break loss-of-coolant accidents of a pressurized water reactor. The inner diameter and the length of a test flow channel used in the experiments were 18 mm and 4 m, respectively. Experiments were performed at higher steam velocity a little than the velocity that was expected just after scram as the first trial. There was no water drainage form the test pipe to the lower plenum. All condensed water was entrained by steam to flow out from the top of the test pipe to the upper plenum. The test pipe was filled with the water lump and the water film, then these were blown up upward and the inner wall of the test pipe became dry. Again the test pipe was filled with the water lump and the water film, then these were blown up upward and the inner wall of the test pipe became dry. This process was iterated at short intervals. The flow state in the test pipe is highly chaotic and agitated. Condensed water flows up and down at high frequencies. It is indicated that to examine the time averaged void fraction and the two-phase pressure drop of the counter-current flow are required.

Commentary by Dr. Valentin Fuster
2014;():V005T17A044. doi:10.1115/ICONE22-30614.

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%.

In terms of a pressure boundary SCWRs are classified into two categories, namely, Pressure Vessel (PV) SCWRs and Pressure Channel (PCh) SCWRs. A generic pressure channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel channel design (for PCh SCWR). Second, a nuclear fuel and fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results.

Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and re-entrant channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant fuel channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles.

A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for fuel and sheath, respectively.

Topics: Pressure , Fuels , Design , Water
Commentary by Dr. Valentin Fuster
2014;():V005T17A045. doi:10.1115/ICONE22-30616.

Pressure drop calculation and temperature profiles associated with fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a fuel channel of a SuperCritical Water-cooled Reactor (SCWR). One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions. The total pressure drop due to all these parameters was between 108 and 121 kPa.

Commentary by Dr. Valentin Fuster
2014;():V005T17A046. doi:10.1115/ICONE22-30652.

The objective of this work is to investigate characteristics of co-current boiling flow in a circular pipe with an inner diameter of 52 mm by using wire mesh tomography (WMT) and ultrasonic velocity profile (UVP). The inner wall of pipe is modified by adding fins on the inner pipe’s wall. This modification is intended to change the flow behavior into swirling flow in boiling flow. Firstly, the effect of wall modification on flow behavior is investigated by numerical calculation. Secondly, two-phase flow is investigated experimentally using UVP and WMT. In experiments, local time-average void fraction is measured using WMT and velocity profile is measured using UVP. Furthermore, these measured data, both void fraction and velocity profile, will give information about changing in flow pattern caused by modified inner pipe’s wall.

Commentary by Dr. Valentin Fuster
2014;():V005T17A047. doi:10.1115/ICONE22-30698.

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.

Commentary by Dr. Valentin Fuster
2014;():V005T17A048. doi:10.1115/ICONE22-30707.

In this work, a Reduced Order Model (ROM) for multi-group time-dependent parametrized reactor spatial kinetics is presented. The Reduced Basis method (built upon a high-fidelity “truth” finite element approximation) has been applied to model the neutronics behavior of a parametrized system composed by a control rod surrounded by fissile material. The neutron kinetics has been described by means of a parametrized multi-group diffusion equation where the height of the control rod (i.e., how much the rod is inserted) plays the role of the varying parameter. In order to model a continuous movement of the rod, a piecewise affine transformation based on subdomain division has been implemented. The proposed ROM is capable to efficiently reproduce the neutron flux distribution allowing to take into account the spatial effects induced by the movement of the control rod with a computational speed-up of 30000 times, with respect to the “truth” model.

Commentary by Dr. Valentin Fuster
2014;():V005T17A049. doi:10.1115/ICONE22-30712.

In this study, validations of Reynolds Averaged Navier-Stokes Simulation (RANS) based on Kenjeres & Hanjalic MHD turbulence model (Int. J. Heat & Fluid Flow, 21, 2000) coupled with the low-Reynolds number k-epsilon model have been conducted with the usage of Direct Numerical Simulation (DNS) database.

DNS database of turbulent channel flow imposed wall-normal magnetic field on, are established in condition of bulk Reynolds number 40000, Hartmann number 24, and Prandtl number 5.

As the results, the Nagano & Shimada model (Trans. JSME series B. 59, 1993) coupled with Kenjeres & Hanjalic MHD turbulence model has the better availability compared with Myong & Kasagi model (Int. Fluid Eng, 109, 1990) in estimation of the heat transfer degradation in MHD turbulent heat transfer.

Commentary by Dr. Valentin Fuster
2014;():V005T17A050. doi:10.1115/ICONE22-30790.

The Advanced High-Temperature Reactor (AHTR) is a new nuclear power reactor concept being investigated in some countries including the United States. The coolant is a liquid salt with a melting point of 460°C and a boiling point of 1430°C. The AHTR uses Silo Cooling System (SCS) as the decay heat removal system in a Beyond-Design-Basis Accident (BDBA). SCS has two accident mitigations. The first component is low-cost, and thick steel rings which conduct heating up the silo wall for BDBA. The second component is an annular ring of an inexpensive, solidified BDBA salt, which is heated from the bottom and melts when the temperature of the salt increases above the melting point, then flows into the silo, and floods the whole silo to its top level. SCS could make AHTR free from catastrophic accidents, where core melting or vessel failure never takes place since the BDBA salt near the top of silo passively absorbs decay heat. On the other hand, AHTR decreases its heat removal ability to avoid freezing of the salt and blocking the flow of the liquid when the temperatures are low. We performed the numerical calculation of AHTR heat removal system and evaluated whether it has the ability to remove decay heat with the robustness for a long-time cooling operation after BDBA. Furthermore, we need to build up and optimize the operation plan of SCS in AHTR, taking its thermal characteristics of this system into account. It is essential to avoid severe accidents which we can suppose as the possible catastrophic scenario.

In this paper, we calculated temperature distributions using the thermal-hydraulics code developed for AHTR, and assessed the performance in a long term cooling period under BDBA conditions. Finally, we investigated the temperature distributions of the whole plant, predicting the accident scenario without air-cooled passive decay heat-removal system. We obtained important conclusion about SCS of the AHTR that its heat removal ability was enough to avoid catastrophic accidents under Loss of Heat Sink (LOHS) conditions.

Commentary by Dr. Valentin Fuster
2014;():V005T17A051. doi:10.1115/ICONE22-30806.

This study was intended to examine sodium entrainment behavior in the case that a hole was formed on a tube wall in the steam generator of a fast breeder reactor and high pressure and high temperature water jetted out into sodium. Flow visualization experiments of an air jet in liquid were performed. The test vessel was 270 mm wide, 5 mm depth and 300 mm high. The air jet was blown vertically upward into stagnant liquid in the test vessel from a rectangular cross-section nozzle of 1 mm wide, 5 mm depth and 20 mm long which was located at the bottom of the test vessel. A flow state of the jet in the liquid was recorded with a high speed video camera at the fastest 150,000 frame/s. The test liquid was water and kerosene. Filament-like ears and wisps pulled out from the wavy interface were noticed on the interface between liquid and the air jet. The ears and the wisps were broken off and entrained into the air jet. The droplets broke up to small entrainments. This process seemed quite similar to the entrainment process in the annular dispersed flow in a pipe. Entrainment was initiated at a little bit downstream from the nozzle outlet. The entrainment inception point moved downstream as the air jet velocity increased. Axial directional entrainment velocity increased as the air jet velocity increased and the entrainment proceeded downstream. Transversal directional entrainment velocity was much slower than the axial directional entrainment velocity. The variation of the entrainment velocity in the transversal direction was not so prominent. The entrainments produced at the interface of the air jet moved to gather at the center portion of the air jet as those were accelerated.

Commentary by Dr. Valentin Fuster
2014;():V005T17A052. doi:10.1115/ICONE22-30812.

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel thermal hydraulics code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in greater detail. Still these code systems lack a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. To our knowledge a two-way coupling to a fuel performance code hasn’t so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models.

A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switching from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and the possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states.

Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up.

The numerical convergence for DYN3D-TRANSURANUS is quick and stable. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.

Commentary by Dr. Valentin Fuster
2014;():V005T17A053. doi:10.1115/ICONE22-30818.

In control-oriented simulators, the neutronics is usually modeled by implementing the point-wise kinetics. In the framework of the study of the control strategy for innovative reactor concepts, such a simplified description is less effective since prevents the possibility of exploiting the capabilities of advanced control schemes. In order to overcome these limitations, a spatial neutronics model based on Modal Method has been considered. This approach allows separating the spatial and time dependence of the neutron flux, which can be represented as sum of the eigenfunctions of the multi-group neutron diffusion equation weighted by time-dependent coefficients. In this way, the system dynamic behavior is reduced to the study of these time-dependent coefficients and can be represented by a set of Ordinary Differential Equations (ODEs), reducing the computational burden. In this paper, a test case involving three fuel pins of an innovative Lead-cooled Fast Reactor has been set up and investigated. Once obtained the eigenfunctions, the set of ODEs for studying the time-dependent coefficients has been implemented in the MATLAB environment [1]. Finally, in order to assess the performance of the developed model, the outcomes have been compared with the results obtained from the neutron diffusion partial differential equation, achieving a satisfactory agreement.

Commentary by Dr. Valentin Fuster
2014;():V005T17A054. doi:10.1115/ICONE22-30821.

A novel, Centrifugally Tensioned Metastable Fluid Detector (CTMFD) sensor technology has been developed over the last decade to demonstrate high selective sensitivity and detection efficiency to various forms of radiation for wide-ranging conditions (e.g., power level, safeguards, security, and health physics) relevant to the nuclear energy industry. The CTMFD operates by tensioning a liquid with centrifugal force to weaken the bonds in the liquid to the point whereby even a femto-scale nuclear particle interactions can break the fluid and cause a detectable vaporization cascade. The operating principle has only peripheral similarity to the superheated bubble chamber based superheated droplet detectors (SDDs); instead, CTMFDs utilize mechanical “tension pressure” instead of thermal superheat offering a lot of practical advantages. CTMFDs have been used to detect a variety of alpha and neutron emitting sources in near real-time. The CTMFD is selectively blind to gamma photons and betas allowing for detection of alphas and neutrons in extreme gamma/beta background environments such as spent fuel reprocessing plants or under full power conditions within an operating nuclear reactor itself. The selective sensitivity allows for differentiation between alpha emitters including the isotopes of Plutonium. Mixtures of Plutonium isotopes have been measured in ratios of 1:1, 2:1, and 3:1 Pu-238:Pu-239 with successful differentiation. Due to the lack of gamma-beta background interference, the CTMFD’s LLD can be effectively reduced to zero and hence, is inherently more sensitive than scintillation based alpha spectrometers or SDDs and has been proven capable to detect below femtogram quantities of Plutonium-238. Plutonium is also easily distinguishable from Neptunium making it easy to measure the Plutonium concentration in the NPEX stream of a UREX reprocessing facility. The CTMFD has been calibrated for alphas from Americium (5.5 MeV) and Curium (∼6 MeV) as well. The CTMFD has furthermore, recently also been used to detect spontaneous and induced fission events which can be differentiated from alpha decay allowing for detection of fissionable material in a mixture of isotopes. This paper discusses these transformational developments which are also being entered for real-world commercial use.

Commentary by Dr. Valentin Fuster
2014;():V005T17A055. doi:10.1115/ICONE22-30845.

Neutron detectors are deployed at ports of entry across the world to monitor people and cargos for smuggled nuclear materials and are often incorporated in nuclear power plant designed to monitor power levels and ensure safe operations. With the supply of Helium-3 rapidly decreasing and due to increase of terrorism threats it is vital to U.S. national security that a viable alternative material be identified, and a new neutron detector design made available, especially for portal monitoring applications. The interest to study Boron-10 as an alternative to helium-3, due to the vast natural supply that the United States possesses and its nontoxic characteristics, is increasing in different research arenas. Our work consists on taking advantage of the near 20% of boron-10 present in naturally occurring boron which (just as phosphorus) is used to dope semiconductors. Boron doped semiconductor wafers were characterized using four-point probe techniques. Resistors and transistors made with various levels of boron concentration were exposed to a thermal flux at various fluencies at Los Alamos National Laboratory. The reaction, 10B+n → Li + α, caused by neutron irradiation, introduces impurities in the silicon lattice thus producing measurable differences in electronic parameters. These changes are likely to be proportional to the fluence of the source, and hence to the neutron flux. The results show that for irradiated resistors possessing very high values of boron concentration there is a significant reduction in resistivity. This trend is not seen for medium or low values of boron. Additionally, there was no observation of significant changes in other electronic parameters such as threshold voltage or trans-conductance, for the transistors exposed and tested.

Commentary by Dr. Valentin Fuster
2014;():V005T17A056. doi:10.1115/ICONE22-30851.

Complex multiphase gas-liquid flows, including boiling, are usually encountered in safety related nuclear applications. For CFD purposes, modeling the transition from low to high void fraction regimes represents a non-trivial challenge due to the increasing complexity of its interface. For example, churn-turbulent and slug flows, which are typically encountered for these gas volume fraction ranges, are dominated by highly deformable bubbles. Multiphase CFD has been so far relying on an averaged Euler-Euler simulation approach to model a wide range of two-phase applications. While this methodology has shown to date demonstrated reasonable results (Montoya et al., 2013), it is evidently highly dependable on the accuracy and validity of the mechanistic models for interfacial forces, which are necessary to recover information lost during the averaging process. Unfortunately existing closures, which have been derived from experimental as well as DNS data, are hardly applicable to high void fraction highly-deformable gas structures. An alternative approach for representing the physics behind the high void fraction phenomena, is to consider a multi-scale method. Based on the structure of the gas-liquid interfaces, different gaseous morphologies should be described by different CFD approaches, such as interface tracking methods for larger than the grid size interfacial-scales, or the averaged Euler-Euler approach for smaller than grid size scales, such as bubbly or droplet flow. A novel concept for considering flow regimes where both, dispersed and continuous interfacial structures, could occur has been developed in the past (Hänsch et al., 2012), and has been further advanced and validated for pipe flows under high void fraction regimes (Montoya et al., 2014) and other relevant cases, such as the dam-break with an obstacle (Hänsch et al., 2013). Still, various short-comings have been shown in this approach associated mostly to the descriptive models utilized to obtain the continuous gas morphology from within the averaged Eulerian simulations. This paper presents improvements on both concepts as well as direct comparison between the two approaches, based on newly obtained experimental data. Both models are based on the bubble populations balance approach known as the inhomogeneous MUltiple SIze Group or MUSIG (Krepper et al., 2008) in order to define an adequate number of bubble size groups with its own velocity fields. The numerical calculations have been performed with the commercially available ANSYS CFX 14.5 software, and the results have been validated using experimental data from the MT-Loop and TOPFLOW facilities from the Helmholtz-Zentrum Dresden-Rossendorf in Germany (Prasser et al., 2007).

Commentary by Dr. Valentin Fuster
2014;():V005T17A057. doi:10.1115/ICONE22-30928.

The use of mechanical or thermal cutting tools in decommissioning of nuclear facilities generates a lot of incandescent particles. Those particles may represent a deterioration risk of the containment barriers associated with a potential risk of fire starting.

The aim of this study is to characterize the incandescent particles emitted by a wheel grinder (in terms of temperature, diameter and velocity) and to follow those parameters all along their path from their emission point to their impact on the air filter. The characteristics of particles correlated with a possible loss of filter efficiency should highlight the destructing particles for the filter. All the measurement techniques used to characterize experimentally the incandescent particles are presented in this article. Particles are characterized in terms of diameter by microscope visualizations. The particle velocity is measured with a high speed camera using Particle Tracking Velocimetry (PTV) technique. An adaptation of a commercial monochromatic pyrometer is achieved to measure the in-flight particles temperature in our specific configuration. All of these techniques have been implemented on an experimental facility which reproduces representative conditions of the cutting processes realized during dismantling operations. Concerning the investigation of the filter, a global and a local approaches about filter degradation are used. The decontamination factor of High Efficiency Particle Air (HEPA) filter is measured, and detailed visualizations of the filter fibers deteriorations are obtained using Scanning Electrons Microscope (SEM).

Commentary by Dr. Valentin Fuster
2014;():V005T17A058. doi:10.1115/ICONE22-30950.

As an effort to enhance the accuracy in simulating the operations of research reactors, a fuel management code system REFT was developed. Because of the possible complex assembly geometry and the core configuration of research reactors, the code system employed HELIOS in the lattice calculation to describe arbitrary 2D geometry, and used the 3D triangular nodal SN method transport solver, DNTR, to model unstructured geometry in the core analysis. Flux reconstruction with the least square method and micro depletion model for specific isotopes were incorporated in the code. At the same time, to make it more user friendly, a graphical user interface was also developed for REFT.

In the analysis of the research reactors, the calculations involving the control rod movement are encountered frequently. The modeling of the control rods differential worth behavior is important in that the movement of the control rod may introduce variations on the reactivity. To handle the problem two effective ways of alleviating the control rod cusping effect are recently proposed, based on the established code system. The methodologies along with their application and validation will be discussed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A059. doi:10.1115/ICONE22-30970.

In order to ensure safety of nuclear installations, thermohydraulics has developed many ways how to predict the behavior of coolant in a heated boiling channel. Accuracy of these predictions can be improved using three-dimensional Computational Fluid Dynamics (CFD) method, which is based on first principles of fluid mechanics. Even though when using CFD, there is a struggle between the accuracy and low computation costs, in many cases CFD can provide feasible improvement of accuracy compared to more traditional approaches. In this research, the focus is set on channel boiling problems, especially those associated with boiling transitions. The phenomenon of critical heat flux (CHF) is investigated using two-phase CFD computation and is compared to experimental data. There is also comparison with other computation methods. When experiment provides some set of data, CFD calculation provides description of the whole flow behavior that provides significantly more information and is of great value during the design process when it gives the understanding of undergoing effects. Besides CHF, general ability of CFD to predict changes in boiling patterns in two-phase channel boiling flows is discussed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A060. doi:10.1115/ICONE22-30979.

This paper presents a comparison of MELCOR calculated results with experimental data for the QUENCH-16 experiment. The analysis for the air ingress experiment QUENCH-16 has been performed by INRNE. The calculations have been performed with MELCOR code. The QUENCH-16 experiment has been performed on 27-th of July 2011 in the frame of the EC-supported LACOMECO program. The experiments have focused on air ingress investigation into an overheated core following earlier partial oxidation in steam. QUENCH-16 has been performed with limited pre-oxidation and low air flow rate. One of the main objectives of QUENCH-16 was to examine the interaction between nitrogen and oxidized cladding during a prolonged period of oxygen starvation. The bundle is made from 20 heated fuel rod simulators arranged in two concentric rings and one unheated central fuel rod simulator, each about 2.5 m long. The tungsten heaters were surrounded by annular ZrO2 pellets to simulate the UO2 fuel. The geometry and most other bundle components are prototypical for Western-type PWRs. To improve the obtained results it has been made a series of calculations to select an appropriate initial temperature of the oxidation of the fuel bundle and modified correlation oxidation of Zircaloy with MELCOR computer code. The compared results have shown good agreement of calculated hydrogen and oxygen starvation in comparison with test data.

Topics: Computers
Commentary by Dr. Valentin Fuster
2014;():V005T17A061. doi:10.1115/ICONE22-31026.

Molten Salt Reactor (MSR) designs are frequently accompanied by a blanket salt. This way the irradiation of the outer reactor wall will be strongly reduced. On the other hand, the barrier between the core and blanket will undergo higher irradiation and it will be necessary to replace it several times during the reactor lifetime. Furthermore, this blanket salt will also have a positive impact on neutron economy by improving the breeding performance.

In this paper a blanket of a generic two fluid molten salt reactor utilizing fast thorium-uranium cycle was investigated. This was done by tracking the evolution of uranium, neptunium and plutonium isotopes with burnup, which was then influenced by removal of uranium from the blanket. A significant reduction in the production of minor actinides was observed.

The uranium vector removed from the core was then investigated for proliferation resistance, using NUREC proliferation resistance metric and comparison with other weapon designs. The evaluation concluded that while the presence of U-232 increases radiological hazard associated with this uranium, thereby erecting a radiological barrier, it cannot be treated as “self-protecting” based on IAEA and NRC standards, requiring 1 Sv/h at 1m dose rate. Moreover ideas on how an interested party could reduce this radiological hazard were discussed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A062. doi:10.1115/ICONE22-31027.

Continually increasing requirements on nowadays full scope PSA L1 and L2 as whole, which is multiplied by importance of specific data for all modes of operation of nuclear power plant, highlight role of input data used in PSA quantification process. This fact also emphasizes the role of capability to process all necessary information to analyze all nuclear plant modes by appropriate way. Even if abovementioned aspects are relevant for all parts of nowadays PSAs, their importance is critical for internal hazards including specific fire analysis. Because internal fire analysis forms one of the most challenging PSA tasks, requiring interdisciplinary work including processing and integration of extensive amount of data in such a way that fire analysis results are fully consistent with internal PSA events and can be directly incorporated into PSA project. Application of tailored information system forms one of the ways to speed up analyzing process, enhances manageability and maintainability of particular PSA projects and provides effective reporting mean to document process of work as well as traceable and human readable documentation for customers. Such information system also allows implementing rapid changes in processing input data and reduces the risk of human error. Usage of information systems for modification of input data for Living PSA is invaluable. Transparent highly automatized processing of input data allows the analyst to obtain more accurate and better insight to evaluate aspects of particular fire and its consequences. This paper provides brief overview of VUJE approach and experience in this area. The paper introduces general purpose of database developed for PSA needs containing data for relevant PSA structure system and components as well as information relevant for flood and fire analyses. Paper explains as this basic data source is enhanced by adding several relatively independent tiers to employ all common data for fire PSA purpose. Paper also briefly introduces capability of such system to generate integrated documentation covering all stages of fire analyses, covering all screening stages of fire analysis as well as future plans to enhance this part of work in such a way to be capable to build automatic interface between PSA model and fire database to enable PSA model parameters automatic updating and expansion of fires in combinations of initiating events (for example Fire and seismic event).

Commentary by Dr. Valentin Fuster
2014;():V005T17A063. doi:10.1115/ICONE22-31050.

This article illustrates the influence of heterogeneity in an infinite lattice of a Molten Salt Reactor moderated by graphite. For a complete description of heterogeneity in a 2D lattice, two variables are needed; in this study the salt share in the unit cell and the channel radius are used. The equilibrium Thorium-based closed-cycle fuel composition is systematically derived for each chosen combination of points, and results such as kinf and the actinide vector composition are calculated. Results show that the heterogeneity effect can indeed be important for optimization of the core design of moderated molten salt reactors.

Commentary by Dr. Valentin Fuster
2014;():V005T17A064. doi:10.1115/ICONE22-31156.

Markov models (MM) are widely used in dependability assessment of complex safety-critical systems, such as NPP I&C system. The main computational difficulties in case of using MM are model size and stiffness, which pose a problems in its construction, storage and solution. Selection of the solution approach and method, based on analysis of such MM features as stiffness and complexity, increases the assessment accuracy. Result of such analysis helps in making decision between direct and indirect research techniques and set of software packages (SP) to provide the high accurate assessment results.

This paper presents the case study for safety assessment of NPP I&Cs. This is a two-channel FPGA-based Reactor Trip System with three parallel tracks on voting logic “2-out-of-3” in each channel. Several solution techniques and SPs were used to analyze and describe the ways in which main modelling risks can be avoid. Analysis of case study results using different SP allows to formulate few application problems: importance of usability-oriented SP selection in case of solving complex MM; achieving an accurate result for stiff MM; support the results verification to ensure the needed level of confidence.

Topics: Chain , Modeling
Commentary by Dr. Valentin Fuster
2014;():V005T17A065. doi:10.1115/ICONE22-31160.

Accident at Fukushima Dai-Ichi nuclear power plant significantly affected the nuclear industry at time when everybody was expecting the so called nuclear renaissance. There is no question that the accident has at least slowed it down. Research into this accident is taking place all over the world. In this paper we present the findings of research on Fukushima nuclear power plant accident in relation to the Czech Republic. The paper focuses on the analysis of human performance during the accident. Lessons learned from the accident and main human errors are presented. First the brief factors affecting the human performance are discussed. They are followed by the short description of activities on units 1–3. The key human errors in the accident mitigation are then identified. On unit 1 the main error is wrong understanding and operation of isolation condenser. On unit 2 the main errors were unsuccessful depressurization with subsequent delay of coolant injection. On unit 3 the main error is the shutdown of high pressure cooling injection system without first confirming that different means of cooling are available. These errors lead to fuel damage. On unit 1 the fuel damage was probably impossible to prevent, however on unit 2 and 3 it could be probably prevented. The lessons learned for the Czech Republic were presented. They can be summarizes as follows: be sure that plant personnel can and knows how to monitor and operate the crucial plant components, be sure that the procedures on how to fulfill the critical safety functions are available in the symptomatic manner for situations when there is no power available at the plant, train personnel for these situations and have sufficient human resource available for these situations.

Commentary by Dr. Valentin Fuster
2014;():V005T17A066. doi:10.1115/ICONE22-31185.

The availability of IVR (In-vessel Retention of Molten Core Debris) strategy at severe reactor accident depend upon the capacity of ERVC (External Reactor Vessel Cooling), i.e. the CHF (Critical Heat Flux) of the external reactor vessel should be higher than the related location heat flux. In this paper, an analysis model of CHF on the downward facing curved surface for pool boiling has been proposed, which adopts the Helmholtz instability analysis of vapor-liquid interface of the vapor jets which penetrating in the thin liquid film underneath the elongated bubble adhering to the lower head outer surface. When the heat flux closing to the CHF point, the vapor-liquid interface becomes highly distorted which resulted in obvious vapor blanket, it will block liquid to feed the thin liquid film underneath the vapor blanket from the bulk region. As a result, the thin liquid film will dry out gradually. As a consequence, the CHF occurs. Based upon this model, spatial variation of CHF about the downward facing curved surface in different subcooling are obtained, and the safety margin of IVR strategy for AP1000 increase with the increase of the subcooling. However, the IVR strategy may be invalid by comparing the CHF with the related local heat flux under the condition of saturated.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2014;():V005T17A067. doi:10.1115/ICONE22-31188.

Safety and security aspects are of meaningful importance in the design of nuclear facilities.

In this study, the attention is so focused on the potential damaging effects that a large civilian airplane impact could bring in safety relevant structures, like a superficial repository similar to El Cabril one.

Safety performances of such a type of superficial disposal facility, subjected to the aircraft impact and fuel burning, have been analysed and discussed.

Conservative assumptions have been made: normal impact on the lateral repository surface, fire scenario based on the amount of fuel burnt.

Load functions (calculated with the Riera approach) and the maximum temperature reached by fuel during its combustion were used as input (boundary condition) in the numerical simulations as well as the damaging phenomena occurring in the concrete structure.

Numerical analyses, by MARC© code, allowed to simulate the thermo-structural performances of the superficial repository.

The obtained results showed that a repository wall thickness, ranging from 0.6 to 0.9 m, is not sufficient to prevent the penetration of wall itself. Despite the ongoing concrete degradation phenomena, the global strength of the repository seemed to be guaranteed.

Commentary by Dr. Valentin Fuster
2014;():V005T17A068. doi:10.1115/ICONE22-31201.

In the transportation of radioactive waste, the package is designed as the major engineered system capable to ensure the containment and provide safety functions, such as radiation shielding, structural integrity against external mechanical and thermal loads, dissipation of the decay heat, etc. Packaging systems are designed in accordance to rigorous acceptance requirements, like the International Atomic Energy Agency (IAEA) ones, so to provide protection to human being and environment against radiation exposure and contamination, particularly in reference accident scenarios including, as it is widely known in literature, drop, puncture, fire and submersion tests. The scope of the present study is to evaluate the structural response and performance in a free drop test condition of a new Italian packaging system that should be used for the transportation of low and intermediate level radioactive wastes. For this purpose the carried out numerical analyses are presented and discussed. The numerical analyses, performed by the finite element MARC® code, simulate the behaviour of the packaging system components: the overpack, gasket, cover lid, bolts and a concrete matrix representative of the radioactive content.

The obtained results for 1.2 m horizontal drop, on a flat and unyielding surface, were critically analysed and also compared to the experimental ones obtained from the experimental test campaign performed at the Unipi test facility on the new Italian packaging system considered.

The stress and acceleration values indicate that the package, although rather local deformations in correspondence of bolts and secondary lid, is capable to withstand the dynamic loading generated during the drop test without any unacceptable loss of the safety features.

Topics: Packaging
Commentary by Dr. Valentin Fuster
2014;():V005T17A069. doi:10.1115/ICONE22-31233.

During a Loss of Coolant Accident (LOCA), the high energy jet from the break may impinge on surrounding surfaces and materials, producing a relatively large amount of fibrous debris (mostly insulation materials). The debris may be transported through the reactor containment and reach the sump strainers. Accumulation of such debris on the strainers’ surface can cause a loss of Net Positive Suction Head (NPSH) and negatively affect the Emergency Core Cooling System (ECCS) capabilities. The U.S. Nuclear Regulatory Commission (U.S.NRC) initiated the Generic Safety Issue (GSI) 191 to understand the physical phenomena involved in this type of event, and help develop the tools to prove the safety and reliability of the existing Light Water Reactors (LWR) under these conditions. Some nuclear power plants have already adopted countermeasures in an attempt to limit the effect of the debris accumulation on the ECCS performance, by replacing or modifying the existing strainer configurations. In this paper, two different strainer designs have been considered and sensitivity analysis was conducted to study the effect of the approach velocity on the pressure drop at the strainer caused by the debris accumulation. The development of the fibrous beds was visually recorded in order to correlate the head loss, the approach velocity, and the thickness of the fibrous bed. The experimental results were compared to semi-empirical models and theoretical models proposed by previous researchers.

Topics: Containment
Commentary by Dr. Valentin Fuster
2014;():V005T17A070. doi:10.1115/ICONE22-31260.

Great operational challenges are placed on nuclear power plants. These challenges are usually reflected in the expansion of fuel cycle length, long-time operation or power uprates. The one way is to optimize the equipment or replace it with equipment with higher efficiency. The second way is to optimize the fuel and its cladding. In this area it is possible to work mainly on the development of new materials which have better nuclear or mechanical properties.

Nuclear power industry is a conservative one. It is necessary to have a detailed knowledge of materials properties of used equipment. Knowledge of the materials behavior is particularly required in the environment where the materials are exposed to neutron flux. This article focuses on new promising materials that can be used in a nuclear fuel, a nuclear reactor or its closest vicinity. Carbon nano materials can be included among these types of materials. Composite materials have generally improved mechanical and thermal properties with addition of nanoparticles. However the additives itself have an impact on the behavior of the neutron field.

This article describes an experiment that examined the behavior of neutrons in carbon nano fibers, carbon nano tubes and nano wires of aluminum oxide. The main goal of the experiment was to determine how neutron scattering is affected, when the sample is exposed to neutron beam. The article presents results, including additional testing of nano materials. Additional tests were carried out to verify the purity and parameters of the investigated samples.

Commentary by Dr. Valentin Fuster
2014;():V005T17A071. doi:10.1115/ICONE22-31262.

The most researched material in nuclear power industry is uranium dioxide however due to strict safety and sanitary restrictions this material can be researched only in specialized research institutes and universities which have sufficient technological background.

For this reason it can be suitable to find material which would show physical properties similar to UO2 but would not suffer by the strict limitations in storage and handling. In this case much more workplaces could be incorporated in the material research and the list of investigated problems could be significantly enlarged.

One of the possible substitutional materials is the cerium dioxide (CeO2) which shows similar chemical and physical properties like UO2 and in some cases shows also similar neutronic properties. The laboratory research was focused on comparing of the basic neutronic properties. For the comparison of nuclear properties the JANIS software [1] was used as it contains cross section libraries of both materials.

It will be shown that similarity of both materials is significant and in several cases application of cerium dioxide as alternative material is possible with sufficient accuracy. As an example of the use of CeO2, research of influence of the SiC content on the reactivity of nuclear fuels is presented in this work.

Commentary by Dr. Valentin Fuster
2014;():V005T17A072. doi:10.1115/ICONE22-31268.

The extraction of dioxouranium(VI) from aqueous nitric acid solutions (3M) into TBP/ionic liquid mixtures (30% v/v), relevant to spent fuel reprocessing, was investigated experimentally in intensified small channel extraction units. The intensified extractor consisted of a “T” shaped inlet and a FEP channel of 0.5 mm internal diameter. Important hydrodynamic characteristics such as, plug length, and plug speed were determined by means of bright field imaging, and were related to the extraction performance of the extraction unit. The amount of dioxouranium(VI) extracted was measured online by using UV-Vis spectroscopy.

Commentary by Dr. Valentin Fuster
2014;():V005T17A073. doi:10.1115/ICONE22-31272.

During a Loss of Coolant Accident (LOCA) a substantial amount of debris may be generated in containment during the blowdown phase. This debris can become a major safety concern since it can potentially impact the Emergency Core Cooling System (ECCS). Debris, produced by the LOCA break flow and transported to the sump, could pass through the filtering systems (debris bed and sump strainer) in the long term cooling phase. If the debris were to sufficiently accumulate at the core inlet region, the core flow could theoretically decrease, affecting the core coolability. Under such conditions, the removal of decay heat would only be possible by coolant flow reaching the core through alternative flow paths, such as the core bypass (baffle). There are certain plant specific features that can play a major role in core cooling from this bypass flow. One of these of key interest is the pressure relief holes. A typical 4-loop Pressurized Water Reactor (PWR) was modeled using RELAP5-3D to simulate the reactor system response during the phases of a large break LOCA and the effectiveness of core cooling under full core blockage was analyzed. The simulation results showed that the presence of alternative flow paths may significantly increase core coolability and prevent cladding temperatures from reaching safety limits, while the lack of LOCA holes may lead to a conservative over-prediction of the cladding temperature.

Commentary by Dr. Valentin Fuster

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