bluuebunny/crossref_metadata_2025 huggingface dataset:
Design and Analysis of the Axial Bypass Compressor Blade of the Supercritical CO2 Gas Turbine
Authors: Ishizuka Takao , Muto Yasushi , Aritomi Masanori , Tsuzuki Nobuyoshi , Kikura Hiroshige
Published: February 2010

A supercritical carbon dioxide (S-CO2) gas turbine can generate power at a high cycle thermal efficiency, even at a modest temperature level of 500–550°C. Its high thermal efficiency is attributed to markedly reduced compressor work at the vicinity of the critical point. Furthermore, the reaction between Na and CO2 is milder than that between H2O and Na. Consequently, a more reliable and economically advantageous power generation system is achieved by coupling with a Na cooled fast reactor. In a typical design, the reactor thermal power, a turbine inlet pressure and an inlet temperature are, respectively, 600 MW, 20 MPa and 527°C. In the S-CO2 gas turbine system, a partial cooling cycle is used to compensate a difference in heat capacity for the high-temperature – low-pressure side and the low-temperature – high-pressure side of the recuperators to achieve high cycle thermal efficiency. The flow is divided into two streams before the precooler. One stream goes to recuperator 2 via a main compressor (MC); the other goes to recuperator 1 via a bypass compressor (BC). The performance and integrity of these two compressors are crucial. As described herein, an aerodynamic design of BC is given. The inlet temperature, inlet pressure, exit pressure and mass flow rate are, respectively, 77°C, 8 MPa, 20 MPa and 1392 kg/s. The salient features of this compressor are its compact size and a large bending stress caused by the large mass flow rate. The number of stages is numerous associated with the large enthalpy rise compared with MC. To achieve as high efficiency as possible, not a centrifugal type but an axial type is examined first. The aerodynamic design was conducted using one-dimensional design method, where the loss model of Cohen et al. is used. Its aerodynamic design enables the use of several stages and provides total adiabatic efficiency of 21 and 87%, respectively. Then, CFD analysis was conducted using “FLUENT”. Blade shapes were prepared based on flow angles and chord length obtained in the aerodynamic design. The CO2 properties in a fluid computer dataset “PROPATH” were used. The features of gas velocity distribution and pressure distribution were confirmed to the fundamental knowledge. The value of the calculated flow rate coincided very well with that of the design.
General Safety Design Approach and Main Safety Features of ATMEA1
Authors: Mathet Eric , Lauret Philippe , Kanagawa Takashi , Castello Ge´rard , Okabe Yasuhiko
Published: February 2010

ATMEA, an AREVA NP and MHI joint venture offers ATMEA1, an innovative and reliable mid-sized evolutionary reactor based on its parent companies’ experience and technology. With a net electrical output between 1100 and 1150 MWe, ATMEA1 is aimed at ensuring competitive power generation. Its operational flexibility and features make it particularly attractive for countries with small or medium size grid. The presentation will focus on and emphasize the general safety design approach and main safety features of this Generation III+ Pressurized Water Reactor (PWR). High level safety, operational objectives, licensability and public acceptance have driven the design: Buildings and all safety systems are being designed in compliance with stringent regulations and worldwide accepted codes and standards, making ATMEA1 a licensable product in any country around the world. Public acceptance is a key element in the nuclear world. Supported by safety features like large commercial airplanes crash protection and resistance to severe accident, ATMEA1 is a robust generation III+ reactor.
Corrosion Inhibition of Carbon Steel in Cooling System Media by Non-Toxic Linear Sodium Octanoate
Authors: Ali A. H. , Abdel Salam O. E. , Waheed A. F. , Abdel-Karim R.
Published: February 2010

Sodium octanoate CH3-(CH2)6-COONa was found to be one of the straight chain aliphatic carboxylates that show good inhibition characteristics towards mild steel. However, as the previously tested concentrations were too high to be applied in the water treatment of secondary circuits, lower concentrations have been investigated in this research. The tested concentrations ranged from 10 to as high as 700 ppm. Potentiodynamic polarization and weight loss tests have been applied at room temperature to study the metal corrosion behaviour in the presence and absence of the tested corrosion inhibitor. The weight loss test simulated the dynamic condition of cooling water as water available at site ‘city water’ was allowed to circulate via pumps through a corrosion test rack in which carbon steel specimens were mounted. Results obtained from both tests were in good agreement. Results revealed that the concentration should not fall below 200 ppm for the tested inhibitor to be effective. The adsorption of inhibitor on the carbon steel surface was found to obey the Langmuir adsorption isotherm model. Thermodynamic parameters of adsorption of the carboxylate have been determined and discussed. The protective film formed due to the presence of the inhibitor at high concentration has been analyzed by Fourier transform infrared (FTIR) spectra. FTIR spectra revealed that octanoate was adsorbed on the steel surface via its functional group.
Fracture Toughness Estimation for Small CT Specimen of Austenitic Stainless Steels Using Stretched Zone Width
Authors: Torimaru Tadahiko , Sugiyama Masanari , Sakamoto Hiroshi , Tanaka Shigeaki , Nakamura Tomomi
Published: February 2010

Austenitic stainless steels in boiling water reactor (BWR) core structures can experience significant fracture toughness reductions at high fluence levels of neutron irradiation. To understand the fracture toughness reductions of irradiated austenitic stainless steels is very important for integrity estimation of core internals. Larger size specimen is needed to obtain the valid JIC of high fracture toughness material. However, it is generally difficult for irradiated material to prepare the enough big specimens for taking valid fracture toughness in ASTM E1820–06 since the material volume is restricted. Stretched zone which is generated between fatigue crack and dimple surface on the CT specimen fracture surface is said to have good linear relationship with JIC. New method for JIC estimation from small CT specimen using stretched zone width was investigated. Cold-worked SUS304 was used for investigation. 0.25–1.5TCT specimens with similar figure of standard dimensions and the CT specimens with different thickness from standard dimensions were prepared for JIC tests. The correlations of specimen width, thickness, critical stretched zone width, JQ and JIC were investigated. JQ obtained from J-R curve tended to decrease with specimen width in the case of similar figure. However, JQ tended to increase with thickness decrease in the case of CT specimen with different thickness from standard dimensions. On the other hand, critical stretched zone width showed the tendency to decrease with thickness decrease in the case of CT specimen with different thickness from standard dimensions. The equation for JIC estimation of invalid small CT specimens was designed by regression analysis using the above relationship. The prediction accuracy achieved within plus-minus 30%.
Void Fraction Measurement of Gas Jet in Sodium Pool
Authors: Nishizaki Masanori , Tsuruoka Hokuto , Sugiyama Ken-ichiro , Narabayashi Tadashi , Ohshima Hiroyuki
Published: February 2010

The occurrence of breakage failure of secondary heat transfer tube due to overheating by sodium-water reaction in LMFBR (Liquid-Metal Fast Breeder Reactor) steam generator has been a main difficult problem to establish its public acceptance. We are studying experimentally the phenomena for the breakage failure of secondary heat transfer tube. An experiment using sodium is difficult because sodium is chemically active and nebulous. The experimental probe needs sufficiently long life in sodium-water reaction circumstance. Because the corrosive jet causes the probe wastage, the probe needs some thickness. But the thicker the probe become, the less spatial resolution it has. Prior to measurements of void fraction in sodium-water reaction circumstance, we measured the void fraction in sodium and in water without chemical reaction by the measurement of electrical resistibvity. We verified whether the probe with diameter of 0.5mm has proper spatial resolution to investigate the pattern of the void fraction distribution in sodium pool and in water. The thermocouple with diameter of 0.5mm can be used also as a wire of the probe. The diameter of nozzle was 3.5mm, and the jet of Ar gas was injected upward in sodium or water pool. The temperatures of sodium pool and water pool are about 443K and 293K respectively. The velocity of Ar gas jet at the nozzle exit was changed from 17.3m/s to 129.8m/s. We used two different kinds of probes, stainless steel wires of 0.5mm and 0.1mm diameters respectively, which were covered by silicone rubber tube of 1.0mm and 0.2mm outer diameter as insulating coating respectively. The length of detection part of the probe (L) was changed from 5.0mm to 0.5mm. We measured the void fraction along the jet-center axis. As L decreased, the void fraction always increased at every 20mm distance from the nozzle exit in sodium and in water. When L was below or equal to about 1.0mm, the void fraction did not increase. Compared the void fraction of the probe with diameter of 0.5mm with that of 0.1mm, the void fraction was almost similar within approximately 10% difference both in sodium and in water. Compared the void fraction in water with that in sodium, the distribution pattern of the void fraction in sodium showed a nearly same trend as that in water though the void fraction in sodium was totally lower than that in water. It is suggested that this trend reflects the fact the surface tension of sodium is higher than that of water. As a result, the probe with diameter of 0.5mm has a proper spatial resolution to investigate the pattern of the void fraction distribution in sodium and in water.
Oxidization and Combustion in Liquid Sodium Droplet
Authors: Otake Shiro , Nishimura Masahiro , Sugiyama Ken-ichiro
Published: February 2010

Liquid sodium is used as the coolant of the fast reactor because of its high thermal conductivity. It is superior as thermal medium in spite of high chemical reactivity with oxygen. In many cases, the combustion starts after forming the dendrite oxides on its reaction surface in oxygen atmosphere. This fact indicates that sodium doesn’t ignite and burn easily unless dendrite oxides are formed. The understanding of the formation mechanism of the dendrite oxides helps us to obtain the optimum handling of leftover non-burning sodium after the accident. However, experimental knowledge to understand the mechanism of combustion is still insufficient. The purpose of this study is to clarify the oxidization behavior of a liquid sodium droplet precisely [1]. The reason why we choose the droplet form is derived from the fact that the reaction surface can be easily observed. The experiment was carried out in a glove box filled with argon gas. A single sodium droplet was made at the tip of the nozzle and preheated at the experimental temperature. The oxidization started by supplying the gas mixture of nitrogen and oxygen. The oxygen concentration of the gas mixture and the initial sodium temperature were adopted as experimental parameters, 4% ∼ 20% O2 and 200°C ∼ 500°C respectively. When the gas mixture was supplied, the droplet surface was covered with a white oxide layer. Gradually, the dendrite oxides appeared on the lower side of the sodium droplet, and the aerosol was generated in the vapor phase area. Then, the dendrite oxides on the surface sank in the droplet and the surface became smooth again. Finally the sodium droplet was encompassed by orange flame. The dendrite oxides were found to grow bigger at the first reaction period. It is suggested that the dendrite oxides have a porous structure and the liquid sodium in the droplet is drawn up to their tips by capillary force. Then the sodium oxidization occurs at the tips. The sodium droplet covered by dendrite oxides was also found to be heated up and the dendrite oxides sank into the droplet due to the high oxygen solubility and be finally burned due to the attainment of its ignition temperature as the second reaction period. The heat generation of oxidization as the first reaction period contributed to the combustion phenomena.
Licensing Principles of FPGA-Based NPP I&C Systems
Authors: Siora Alexander , Sklyar Volodymyr , Rozen Yuriy , Vinogradskaya Svetlana , Yastrebenetsky Mikhail
Published: February 2010

This paper objective is to discuss licensing principles of FPGA-based NPP I&C systems. Licensing aspects of FPGA-based NPP I&C systems lay in peculiarities of FPGA-technology. The main idea for licensing performing of FPGA-based NPP I&C systems lays in consideration of FPGA-chip as hardware and FPGA-project as software. FPGA related licensing principles are discussed in details in the paper.
Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility
Authors: Parfenov Yuriy V. , Melikhov Oleg I. , Melikhov Vladimir I. , Elkin Ilya V.
Published: February 2010

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.
Air Entrainment by Impinging Jets: Experimental Identification of the Key Phenomena and Approaches for Their Simulation in CFD
Authors: Schmidtke Martin , Danciu Dana , Lucas Dirk
Published: February 2010

The gas entrainment by impinging water jets is studied by the means of experiments and CFD simulations. In the experiments, the nozzle height above the water pool and the water velocity inside the nozzle is varied and the bubble entrainment below the jet is observed with a video camera. Quantitative data for bubble sizes and the geometry of the bubble plume are obtained by image processing. The development of two phase jets under water is also studied with CFD, using the Euler-Eulerian approach. Defining a criterion for an average penetration depth, the simulated plumes are compared with those observed in the experiments.
Evaluation of the Inter Granular Stress Corrosion Cracking Defects in Austenitic Stainless Steel Piping
Authors: Janulionis Remigijus , Dundulis Gintautas , Karalevicˇius Renatas
Published: February 2010

The Inter Granular Stress Corrosion Cracking (IGSCC) is a dominant damage mechanism of the austenitic stainless steel. The primary circuit piping of RBMK type reactors is produced from austenitic stainless steel 08×18H10T. Defects in welded joints of pipes with nominal diameter of 300 mm were detected during In-service inspections [1]. Metallographic investigations defined that crack growth mechanism is IGSCC. The appearance of defects increases the probability of RCS piping failures of these pipes. A leak or break in RCS piping is not acceptable from safety and political (society risk) points of view. According this the evaluation of these cracks is very important for safe operation of this type reactor. The procedures for IGSCC crack evaluation consist of two parts. The first part is determination of the acceptable crack size for the component with crack, and the second part is the crack growth calculation. The acceptable flaw size provides information about the largest flaw size which component can tolerate without failure with accepted safety factors. The crack growth calculation determines how long does it take for the existing crack to reach the maximal acceptable size. The results of these calculations (acceptable crack size and crack growth) determine the further inspection schedule of the components with crack. The objective of this paper is the evaluation of the IGSCC defects detected during In-Service inspection in the primary circuit piping which outside diameter of piping is 325 mm, the wall thickness – 16 mm. Detected cracks were evaluated using method R6 [2]. The IGSCC crack growing analysis was performed using methodology presented in document [3]. The prognosis results were compared with crack data detected during In-service inspection. According analysis results were determined that the IGSCC defects detected during In-service inspection can be left without repairing for 1.5 years operation.
Numerical Analysis of Supercritical Water Heat Transfer in Horizontal Circular Tube
Authors: Zhang Bo , Shan Jianqiang , Jiang Jing
Published: February 2010

CANDU supercritical water reactor (SCWR) offers advantages in the areas of sustainability, economics, safety and reliability and proliferation resistance. However, there is still a big deficiency in understanding and prediction of heat transfer behaviour in supercritical fluids. In this paper, heat transfer is numerically investigated on supercritical water for three-dimensional horizontal flows. Three ε-type turbulence models are tested and the numerical results are compared with experimental data. Based on the result, the standard k-ε turbulence model with enhanced wall treatment is recommended. The effect of the buoyancy and heat transfer deterioration is also analyzed, and the criteria for onset of buoyancy effects is evaluated. The quantity Gr/Re2.7 recommended by Jackson et al. (1975) gives a capacity to predict the buoyancy.
AP1000™ Breaking the Dependence on Site Characteristics: Electrical Grid and Cooling Water
Authors: Anderson Richard G. , Schulz Terry L. , McLaughlin Daniel T.
Published: February 2010

The AP1000™ has been developed to use passive safety systems to address design basis and beyond design basis accidents to minimize impact on calculated plant Core Damage Frequency (CDF) and Large Release Frequency (LRF). Passive safety cases are supplemented with accident mitigation strategy using highly reliable non-safety grade systems. AP1000™ non-safety systems that resemble the safety systems of conventional plants are designed to mitigate accidents, when available. In addition, a number of mitigation schemes make use of non-safety systems together with select passive system features. This design approach has resulted in a small dependence on site characteristics and has minimized their contribution to CDF and LRF. Conventional nuclear power plants are designed to use motor driven safety grade equipment to address design basis accidents. Safety grade diesel generators are used to provide power to safety grade equipment in the event of a loss of offsite power. The contribution to CDF and LRF from Loss of Offsite Power (LOSP) or loss of cooling water events is significantly lower for the AP1000™ than that for conventional plants. The AP1000™ uses passive safety systems in conjunction with non-safety systems to reduce plant CDF and LRF sensitivity to site specific characteristics. Non-safety systems used for accident mitigation make use of high grade commercial components and are provided power from non-safety grade diesel generators upon LOSP. These components are well maintained to increase system reliability and to increase availability for accident mitigation. The effect of the non-safety systems is further enhanced by partial use of features of the passive safety systems for accident mitigation. In this way, events with initiating event frequencies often driven by site characteristics are mitigated with small contribution to CDF or LRF, often without the need to activate any, or some of the passive plant features.
Linear and Non-Linear Stability Analysis of a Supercritical Natural Circulation Loop
Authors: Sharma Manish , Vijayan P. K. , Pilkhwal D. S. , Saha D. , Sinha R. K.
Published: February 2010

Supercritical water has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN has been developed employing supercritical water properties to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearised by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been benchmarked against published results. Then the code has been extensively used for studying the effect of diameter, heater inlet temperature and pressure on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). A separate computer code NOLSTA has been developed which investigates stability characteristics of supercritical natural circulation loop using non-linear analysis. The conservation equations of mass, momentum and energy in transient form were solved numerically using finite volume method. The stable, unstable and neutrally stable points were identified by examining the amplitude of flow and temperature oscillations with time for a given set of operating conditions. The stability behavior of loop, predicted using non-linear analysis has been compared with that obtained from linear analysis. The results show that the stability maps obtained by the two methods agree qualitatively. The present paper describes the linear and nonlinear stability analysis models and the results obtained in detail.
Structural and Neutronic Assessment of Calandria Tube Enhancement Design Options
Authors: Scrannage Richard C. , Sani Farshad , Hobbs Daniel M. , Freeburn Allan M. , Harvel Glenn D.
Published: February 2010

Over the lifetime of a CANada Deuterium Uranium (CANDU) Type Reactor, the pressure tubes and calandria tubes deform via creep and sag, eventually requiring replacement. The sagging phenomena leads to fuelling issues, potential contact between the Calandria Tube (CT) and the liquid injection shutdown system or the Pressure Tube (PT). As such, this ageing phenomenon limits the lifetime of the components. Conceptual designs were assessed to determine their potential for reducing the effects of aging by minimizing the magnitude of sagging. Three different design options for a CT have been investigated using numerical simulation techniques. The CT design options include fins and ribs of different sizes and combinations. The fins and ribs provide improved structural integrity over the reactors lifetime. Analyses results have shown that the design options yield an increased overall strength with a minimal impact on neutronics. The analysis has determined that the finned design option is superior in terms of CT strength enhancement and minimizing the impact on neutronics. Furthermore moderator displacement analysis for the finned design option showed an ability to minimize the impact on moderation or even cause moderator displacement potentially decreasing the positive void coefficient.
Advanced Recycling Core Accommodating Oxide Fuel and Metal Fuel for Closed Fuel Cycle
Authors: Ikeda Kazumi , Maddox James W. , Nakazato Wataru , Kunishima Shigeru
Published: February 2010

This report presents a unique core capable of accommodating oxide fuel and metal fuel and easy to change oxide core to metal core conforming to the design requirements. For the homogeneous oxide fueled core containing TRU (transuranics) fuel with 12% of the moderator pins, the results of calculation show the TRU conversion ratio (ratio of loss of TRU to loss of heavy metal) of 0.33 and the TRU burning capability (ratio of loss of TRU per electric generation) of 67 kg/TWeh. On the other hand, the calculations replacing from oxide fuel assemblies to metal fuel assemblies have demonstrated the TRU transmutation capability of 69 kg/TWeh with the TRU conversion ratio of 0.30. As the result of simulation calculations, three ordinary fuel exchanges the oxide equilibrium core to full metal core by way of transitional cores, where the maximum linear heat rates are still equal to the metal equilibrium core or less. With this, the presented core concept is concluded that a full oxide core, a full metal core, mixed fueled cores can be materialized in the presented ARR’s first unit (ARR1).
Designing Mass Exchangers for Control of Oxygen Content in Pb-Bi (Pb) Coolants in Various Research Facilities
Authors: Martynov P. N. , Askhadullin R. Sh. , Simakov A. A. , Chaban’ A. Yu. , Legkikh A. Yu.
Published: February 2010

The solid-phase method designed by the State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering (SSC RF - IPPE) for the oxygen thermodynamic activity control [1, 2] implies the use of oxygen source, i.e. dissolvable lead oxide spheroids hold within limited section of the circuit (reaction vessel of mass exchanger) connected by the pipeline to the main circuit. PbO spheroids being in contact with flowing lead-bismuth (lead) are dissolved with generation of oxygen, which is transported throughout the circuit by the coolant flow. Technological implementation of solid phase control method is based on the use of specially designed mass exchangers (MX) being an important component of heavy liquid metal coolant technology.
Influence of Impurities on Intergranular Corrosion of Extra High Purity Austenitic Stainless Steels
Authors: Ioka Ikuo , Suzuki Jun , Motoka Takafumi , Kiuchi Kiyoshi , Nakayama Junpei
Published: February 2010

An intergranular corrosion is observed in austenitic stainless steels exposed to high temperature, concentrated nitric acid (HNO3) solution with highly oxidizing ions. It is an important degradation mechanism of austenitic stainless steels for use in a nuclear fuel reprocessing plant. The intergranular corrosion is caused by the segregation of impurities to grain boundaries and the resultant formation of active sites. Extra High Purity (EHP™) austenitic stainless steel was developed with conducting the new multiple refined melting in order to suppress the total harmful impurities less than 100ppm. The intergranular corrosion behavior of EHP alloys with various impurities was examined in boiling HNO3 solution with highly oxidizing ions to find a correlation between the intergranular corrosion and the impurities of EHP alloys. A good correlation was confirmed between the degree of intergranular corrosion and the corrosion rate. The relationships between the corrosion rate and the impurities content of EHP alloys was determined using a multiple regression analysis. The influence on corrosion rate became small in order of B, P, Si, C, S and Mn. It was important to control B in intergranular corrosion behavior of EHP alloys.
Use of BEPU Technique for Analyses of BDBAs With Cooling in WWER-1000
Authors: Ponomarenko Grigory L. , Ryzhov Sergey B. , Bykov Mikhail A. , Moskalev Aleksandr M.
Published: February 2010

Hypothetical conditions with deep cooling of coolant are analyzed for serial Reactor Installation (RI) with WWER-1000. The beyond design basis accident (BDBA) scenarios are specially chosen, to reveal the effectiveness of internal properties of nuclear fuel and reactor emergency protection (EP) in conditions when criterial parameters are close to the acceptance criteria or violate them. Various phases of the accident were considered: ATWS (anticipated transients without scram), repeated criticality and ‘cold slug’. Plural failures of Control Rods of Control and Protection System (CRs CPS) were modeled at the operation of EP. The analysis was carried out by the coupled code KORSAR/GP [1–3]. At the first investigation stage the individual conservative variant was considered, from which the expediency of use of BEPU method is revealed. Then at the second investigation stage the analysis of the set of variants in BE approach with uncertainty analysis on the set of statistically varied parameters — neutron-physical, thermo-physical, thermo-hydraulic and hydrodynamic was carried out with PANDA+KORSAR/GP [4] program complex. The advantages of Best Estimate Plus Uncertainty (BEPU) method for the safety margins revealing are demonstrated.
Development of Simulation Tool for Dismantling Planning
Authors: Zhu Haishan , Liu Zhongkun
Published: February 2010

The planning of nuclear reactor’s decommissioning is a time consuming process. Visualization and virtual reality technologies can give an intuitive and innovative way for the dismantling planning. This paper introduces some techniques about the visualization simulation of dismantling planning and gives an example on creating the simulation and animation of a worker’s motions plan. The results show that the use of active objects and the graphic modeling can improve the simulation performance and reduce the simulation modeling works.
New Method for Numerical Prediction of Water Hammer With Column Separation: Comparison With Experiment and the Relap5 Program
Authors: Adamkowski Adam , Lewandowski Mariusz , Marcinkiewicz Jerzy
Published: February 2010

Liquid column separation accompanying water hammer (hydraulic transients) was a subject of numerous research works conducted in many scientific centres of the world. So far, there are not any satisfactory results of using computational methods to simulate this phenomenon. The authors of this paper have developed a new method for calculating pressure fluctuations in pipelines during hydraulic transients with liquid column separation. This method is based on the innovative (original) discrete-vapour-cavity model. In the paper there are two steps presented in the validation process of the new model. The first step is the comparison with experimental data and the second one is the comparison with results obtained using the commercial computer program Relap5. The comparisons of calculated by means of the considered numerical methods and measured pressure surges show significant differences. The reasons for the discrepancies are discussed and conclusions are presented.
Optimizing the Initial Pressure of Accumulators for the Atucha2 NPP Using an Optimization Method
Authors: Muellner N. , Del Nevo A. , Cherubini M. , D’Auria F. , Mazzantini O.
Published: February 2010

Passive safety systems like hydro accumulators offer high reliability and are therefore, when a choice is possible, often preferred over active safety systems. However, their effectiveness in case of an incident or accident depends on a large number of parameters (like break size in case of a loss of coolant accident, availability of other safety systems, initial pressure in the accumulators) and is in general difficult to predict. This paper presents a study to optimize the initial pressure and the pressure drops in the accumulator line for a intermediate break loss of coolant accident for Atucha 2, a Siemens-KWU, heavy water moderated, channel type pressurized water reactor under construction. An optimization method was applied. The thermal hydraulic system code RELAP5 mod 3.3 was used for the analysis. Three cases have been analyzed. First, the initial pressure and pressure drop in the accumulator line was optimized in case of an intermediate break in cold leg two, assuming safety injection of two of the four trains of safety systems into hot and cold leg. Second, like before, but assuming safety injection into cold leg only. Third, like case two, but grouping the four accumulators in two groups, with different initial pressure and pressure drops in the accumulator lines. The results show that a slight increase of the initial accumulator pressure compared to the design value could be beneficial for the investigated initial event. Further, case three shows that different initial pressure in the accumulators could increase the effectiveness of the intervention for the investigated accident.
Parameter Adaptation Techniques for the Best-Estimate Thermal-Hydraulic Code ATHLET
Authors: Kerner Alexander , Schaefer Anselm , Chen Kan
Published: February 2010

Parameter adaptation can be used for automated adjustment of real time simulation according to measured data from the simulated plant. It is able to prevent that model uncertainties lead to an increasing deviation between simulation results and real plant behaviour. The objective is to ensure that the simulation corresponds with reality even over a longer period. The present investigation deals with the application of parameter adaptation techniques to best-estimate simulation. This opens a range of new applications including the use of best-estimate codes for improving process diagnostics, signal validation and the man-machine-interface. The adaptation of ATHLET poses several challenges. Among them are non-linear dependencies within the equation system to be solved, the interference of a sophisticated time integration technique, and the identification and proper consideration of those code uncertainties which are most influential with regard to deviations between simulation and real plant behaviour. To overcome these problems a set of new techniques has been developed: - A refined method for the definition of the parameters to be adapted is based on the consideration of the parameter’s uncertainty and sensitivity to the results. Statistic sensitivity analysis is used to validate and improve parameters obtained by expert judgement. - Different parameter adaptation techniques (e.g. based on standard controllers or parameter estimation techniques) are applied to produce feasible solutions for the large systems to be modeled. - In view of preventing non-reversible propagation of simulation inaccuracies decomposition techniques are applied assuring that such temporary inaccuracies do not spread over subsystem boundaries. Results of adaptation based on these techniques are shown for an experiment performed at the test facility PKL III. It turns out that adaptive simulation is feasible for such facilities and that it offers possibilities to improve prediction of experiments by code calculations.
Modernization of the Safety I&C at the Dukovany VVER 440/213 Nuclear Power Plant
Authors: Palaric Jean-Michel , Rebreyend Philippe , Mouly Philippe , Esmenjaud Claude , Dalik Frantisˇek
Published: February 2010

The modernization of the Dukovany nuclear power plant (four VVER 440 MWe reactor units owned by CˇEZ, the Czech national utility) is presented with a special focus on the digital safety instrumentation and control (I&C) system. The first Unit has been successfully modernized in compliance with the initial schedule. The following matters are further discussed in this paper: • Goal, scope and industrial organization of this modernization, • Main design criteria and I&C architecture, • Digital technologies in use, • Design and Licensing processes, • On-site installation strategy and main milestones, • Progress of work.
CFD Modeling of Fuel Rod Heat Transfer to Aid in the Risk Assessment of Fuel Rod Failure Due to Localized Crud
Authors: Beltz Steven , Liu Bin , Karoutas Zeses
Published: February 2010

This paper presents a computational fluid dynamics (CFD) modeling methodology that has been developed to provide predictions of very local heat transfer variation in fuel rod assemblies. Results from the CFD analysis are used in HIDUTYDRV and other advanced codes that have been developed and are used internally by Westinghouse to predict very local crud deposition and dryout. This methodology is used in making the EPRI Level IV crud and corrosion guideline assessments, which were developed in response to the INPO 0 by 2010 initiatives. This methodology has been in production use for risk assessment of CE-design 14×14 and 16×16 fuel reloads. The methodology is in the process of being extended to other Westinghouse fuel design reloads. Local crud deposition and dryout are strongly dependent on very local boiling or steaming on small areas of the fuel rod, often referred to as local hot spots. These local hot spots can not be predicted utilizing standard subchannel modeling methodology because subchannel models do not provide sufficient azimuthal detail of individual rods. Local hot spots are also very dependent on the particular grid features, which are not explicitly modeled in subchannel analysis. The commercial code Star-CD by CD-ADAPCO is utilized to develop a detailed CFD model of a single fuel assembly grid span. Detailed azimuthal and axial predictions of the heat transfer coefficient are made for each rod in the model. These predictions are then normalized to a Dittus-Boelter based heat transfer coefficient so that the predictions can be translated to other spans and other fuel assemblies. Details of this translation as well as the use of normalized heat transfer coefficients in the advanced codes used to predict local crud and dryout are provided in a separate follow-on paper ICONE17-75715 also being presented at ICONE17. This paper presents details on the CFD methodology that has been developed to predict local normalized heat transfer coefficients for a fuel rod assembly. Results for a particular application are provided to illustrate the methodology. The application is for a fuel design that contains mixing grids and spans with and without intermediate flow mixers.
How to Obtain an ASME Code Nuclear Certificate of Authorization
Authors: Bobo Barry K. , LaRochelle William C.
Published: February 2010

Although the procedure for obtaining an American Society of Mechanical Engineers (ASME) Code Nuclear Certificate of Authorization has changed little since 1974, much of the process is not fully understood by designers and manufacturers of pressure equipment. It is the intent of this paper to make this process easier to understand and explain the key process steps in order to avoid unnecessary delays. The discussion will cover a number of topics including: “Contracting with an Authorized Inspection Agency,” “Deciding the Type of Certificate to Pursue. “Quality Assurance Program Preparation,” “Design Considerations,” “Preparing Welding and Non-Destructive Examination Documents,” “Training,” and “The Survey.” The requirements for obtaining one of the various ASME Certificates of Authorization are somewhat different for each. This paper focuses on an N-type Certificate and is intended to be a high level examination of the requirements and typical pitfalls in obtaining that certificate.
Three Dimensional Modelling of the Hydrodynamics of Oblique Droplet-Hot Wall Interactions During the Reflood Phase After a LOCA
Authors: Chatzikyriakou D. , Walker S. P. , Belhouachi B. , Narayanan C. , Lakehal D. , Hewitt G. F.
Published: February 2010

During the reflood phase, following a LOCA, the main mechanism for the precursory cooling of the fuel is by convective heat transfer to the vapour, with the vapour being cooled by the evaporation of the entrained saturated droplets. However, it is believed that the droplets that reach the rod could have an effect on this cooling process. Despite the fact that those droplets do not actually wet the fuel rod due to the formation of a vapour film that sustains them and prevents them from touching the wall, the temperature drop caused by the impingement of such water droplets on a very hot solid surface (whose temperature is beyond the Leidenfrost temperature [1]) is of the order of 30–150 C, [2],[3]. The associated heat flux is of the order of 105–107W/m2 and the heat extracted is in the range of 0.05 J over the time period of the interaction (a few ms) [2],[3]. The hydrodynamic behaviour of the droplets upon impingement is reported to affect the heat transfer effectiveness of the droplets. In the dispersed flow regime the droplets are more likely to impinge on the hot surface at a very small angle sliding along the solid wall, still without actually touching it, and remaining in a close proximity for a much larger time period. This changes the heat transfer behaviour of the droplet. Here, we investigate numerically the hydrodynamics of the impingement of such droplets on a hot solid surface at various incident angles and various velocities of approach. For our simulations we use a CFD, finite volume computational algorithm (TransAT©). The Level Set method is used for the tracking of the interface. We present three-dimensional results of those impinging droplets. Validation of our simulation is done against experimental data already available in the literature. Then, we compare the findings of those results with previous correlations.
Implementation of RAVE™ (To a Reload)
Authors: Gu¨ler Cenk , Lider Serhat
Published: February 2010

Westinghouse 3D Transient Neutronics methodology, RAVE™, has been implemented for the first time for licensing basis analysis of a Westinghouse 2-Loop type Pressurized Water Reactor (PWR). Westinghouse is using the traditional reload methodology for its fleet by utilizing a bounding approach. Safety limits are defined at the beginning of the process by using bounding values. These limits are then checked for every reload and they are shown then there are no violations to those limits therefore the safety calculations are still acceptable for the new loading pattern. This approach eliminates repeating safety/accident calculations for each reload. A new reload methodology was developed which utilizes similar safety reload methodology.
Study on Subcooled Flow Boiling Heat Transfer Under Vibration Conditions
Authors: Ohtake Hiroyasu , Uchida Naoki , Koizumi Yasuo
Published: February 2010

Onset of nucleate boiling — ONB — and critical heat flux — CHF — on subcooled flow boiling under vibration condition, focusing on liquid velocity, amplitude and frequency of vibration, were investigated experimentally. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. The test section was set on a vibration table. The acceleration was 1.3 and 4.2 m/s2, respectively; the frequency was 2, 4 and 20 Hz, respectively. The test section was arranged for horizontal position facing upward and for vertical position, respectively: the vibration directions were horizontal and vertical to orientation of the heater. The present experimental results showed that critical heat fluxes under the vibration conditions of both vibration directions, i.e., horizontal and vertical vibrations to the heater, were higher than those for steady flow. The CHF under the vibration condition was increased with an increasing of acceleration of vibration. According to present observations, coalesced bubble on the heater was frequently released by vibration of the test heater. This behavior causes the CHF to become higher under the vibration condition.
Cost Advantages of Large Underground Nuclear Parks
Authors: Giraud Kellen M. , Kunze Jay F. , Mahar James M.
Published: February 2010

Underground nuclear power plant parks have been projected to be economically feasible compared to above ground installations. This paper is a conceptual analysis of the cost savings, compared to surface facilities, which result from reduced costs in the aspects of construction, transportation of materials, security, and decommissioning. The paper also explores the cost burdens associated with underground nuclear power plant parks. Overall, the cost savings are projected to far outweigh the cost burdens if design and regulatory issues are reasonably managed. The cost savings for electricity generated over a 60 year life of a typical 1000 MWe nuclear power plant are projected to be 0.23 to 0.66 cents (2008 U.S. currency) per kWh.
Numerical Analysis for Heat and Mass Transfer of Granular Flow in a Duct by the Discrete Particle Simulation
Authors: Yamaguchi Tomohiko , Kanemaru Kuniyasu , Momoki Satoru , Shigechi Toru , Fujiwara Ryo
Published: February 2010

The solid-gas or liquid-gas two phase flow has many industrial applications such as spray drying, pollution control, transport systems, fluidized beds, energy conversion and propulsion, material processing, and so on. Though the solid-gas multiphase flow has been studied experimentally and numerically, the transport phenomena have not been cleared due to its complexity, computational time and economical costs for the hardware. In this study the heat and mass transfer of solid-gas collision dominated flow is analyzed by the Discrete Particle Simulation (DPS), a kind of the Dispersed Element Method (DEM)[1]. This method describes the discrete phase and the continuous phase by Lagrange and Euler methods respectively, and has been used to simulate the multiphase flow of various geometrical systems. In order to analyze the thermal field we took account of the energy equation and heat conduction between colliding particles. The heat transfer rate is summation of conductive heat transfer and convective heat transfer. Furthermore, the fluid flow has a two dimensional velocity profile, because the void fractions are analyzed as two dimensions. But momentum space has not been resolved by the two dimensional simulation. We call this method, the quasi two-dimensional simulation in this paper. To obtain the temperature distribution of the continuous phase the energy equation is solved in addition to the momentum equations. We treated the interaction between continuous and discrete phases as one and two way couplings. The positions, the momentum and the temperature information of particles and the velocity and the temperature distribution of the fluid were obtained as functions of time from results of these numerical simulations. When the hot air that is suspending small glass particles flows in a duct from bottom up, we traced the particles and got the temperature distribution of fluid and compared with the former results of one-dimensional flow. At the beginning, the cooler particles decrease the fluid temperature near the bottom of the vessel. The temperature profile of the particles obtained by the one-dimensional simulation is as same as quasi two-dimensional simulation. After 0.5 second the particles cool the downstream air. At 1.2 second, particles do not decrease the air temperature because the temperatures of particles are close to the inlet temperature of the air.
Steady Propagation of the Vaporization Front in Metastable Liquid
Authors: Aktershev Sergey , Ovchinnikov Valeryi
Published: February 2010

The boiling up of a metastable liquid when the vaporization fronts appear is considered by theoretically and experimentally. Boiling up occurs as usually on a surface of a heater. At the first stage the growth of a spherical vapor bubble is observed. If the temperature of liquid exceeds threshold value, the vaporization fronts develop near to line of contact of a vapor bubble and heater. Fronts of vaporization extend along a heater with constant speed. It is direct transition from one-phase convection to a film boiling. Such scenario of crisis of a convective heat transfer also is possible in the nuclear reactor equipment. The model of steady propagation of the vaporization front is developed. The temperature and velocity of propagation of the interface are determined from the balance equations for the mass, momentum, and energy in the neighborhood of the vaporization front and the condition of stability of motion of the interface. It is shown that a solution of these equations exists only if the liquid is heated above a threshold value. The velocity of propagation of vaporization front has threshold value also. The calculated velocity of interface motion and the threshold value of temperature are in reasonable agreement with available experimental data for various liquids within wide ranges of saturation pressures and temperatures of the overheated liquid.