bluuebunny/crossref_metadata_2025 huggingface dataset:
Differential-Discrete Mathematical Model of the Recuperative Counter-Flow Heat Exchanger
Authors: Simeunovic´ G. , Zi´tek P. , Debeljkovic´ D. Lj.
Published: July 2009

The heat exchangers are always constructive elements of nuclear power plants and their dynamics is very important since they introduce a significant time lag in control actions. That’s why their dynamics cannot be ignored. On the basis of the accepted and critically clarified assumptions, a linearized mathematical model of the recuperative counter flow heat exchanger has been derived, taking into account the wall dynamics. The model is based on the fundamental law of energy conservation, covers all heat accumulation storages in the process, and leads to the set of partial differential equations, which solution is not possible in closed form. In order to overcome the solution difficulties, the procedure of differential discrete modelling is applied, leading to the set of ordinary differential equations of a rather high order. These equations are transformed into the state space form suitable for the Matlab environment. The experimental investigations were made, putting some different typical input variables. Specifying the input temperatures and output variables, under the constant initial conditions, the step transient responses have been simulated and presented in graphic form for the particular positions in the heat exchanger in order to compare these results with the experimental data collected from the real process.
Experimental Study of Transient-State Characteristics of Temperature in the Density Lock
Authors: Gu Hai-Feng , Yan Chang-Qi
Published: July 2009

Through the method of visible observation, experimental study on transient-state characteristics of temperature on the interface of stratified fluids is made in three different experimental tubes. On the basis of experimental study, mechanism of temperature oscillation and heat transfer behaviors on the interface of stratified fluids are analyzed under transient-state conditions in the density lock. The results show that the temperature of work fluid near the interface has two kinds of instabilities—high amplitude oscillation and low amplitude oscillation. High amplitude oscillation results from the consequent breaking of the interface and will bring vigorous heat transfer between stratified fluids. After the high amplitude oscillation takes place, temperature jump of work fluid will arise on the interface. However, low amplitude oscillation is only dynamic wave of interface around its balanced position, and it will not cause vigorous heat transfer. Besides, comparative experiment on transient-state characteristics of temperature is done in experimental pipes with different diameters, the results show that high amplitude oscillation is more likely to take place in experimental pipes with larger diameters. Therefore, by virtue of its small diameters, honey-comb channel in density lock can curb the occurrence of high amplitude oscillation, efficiently decreasing heat transfer between stratified fluids.
A Real-Time Interactive Evaluation Method on Human Machine Interface Design
Authors: Xu Yuqing , Ma Qingxin , Yang Ming , Yan Shengyuan
Published: July 2009

During the process of information obtaining and operation executing in main control room of nuclear power pant, the displays or control equipments must be designed according to human factor standards. The design quality of human machine interface (HMI) is influenced by many factors, such as the percentile of human models, the position and structure of display or control equipments and so on, because the relationship between these factors are very closed, so it is difficult for designers to use enough human factor parameters in design process. Although the software for human factor evaluation had been developed, they can not be compatible with 3D modeling environment. Therefore, it is impossible to evaluate the design scheme of human machine interface during its developing process. In this paper, a new human machine interface evaluation method based on 3D modeling environment is proposed, which not only can realize the real-time interactive evaluation in the designing process of structure and position parameters of displays and control equipments, but also can improve the evaluation efficiency. Meanwhile, it can ensure the compatibility between evaluation software and 3D modeling environment. Based on the guidelines of the aforementioned theory, a kind of object-oriented evaluation software is developed under Unigraphics. 3D Human model plays a very important role in the evaluation on human machine interface. Nowadays, there are no human models in China which can be directly used in the human factor evaluation on 3D equipment model. The foreign human models can not be used in China directly, because the human dimension of Chinese adult is quite different. In this paper, firstly, the human model with relevant parameters of human factor evaluation is developed, and it can create Chinese human model of any percentile with different gestures. Secondly, the evaluation software embedded with human factor standards is developed, and it can realize real-time interaction between human model and display or control equipments. With the developed evaluation software, the evaluation on displays controls and consoles could be carried out, as well as the comprehensive evaluation on the whole human machine interface. At the end of this paper, the developed software is used to evaluate the human machine interface of main control room of one power plant. According to the quantitative evaluation results, some suggestions are proposed for improving the design quality of human machine interface.
Problems With Cooling of Overheated Fuel Channels in RBMK-Type Reactors
Authors: Uspuras Eugenijus , Kaliatka Algirdas
Published: July 2009

One of the most dangerous beyond design basis accidents for all types of nuclear reactors is the loss of long-term heat removal from the core. In RBMK-type reactors, this initiating event, which can lead to the worst consequences, has significant probability to occur in comparison to other type of BDBA. The most effective accident mitigation measure in this case is “bleed and feed” strategy — similar as is recommended for other light water reactor types. In this paper the challenges, which are meet in case of cooling of overheated fuel channels in RBMK-type reactors, are discussed. The simulation results of BDBA using RELAP5/MOD3.3 code are presented. Accident management measures (de-pressurization of reactor cooling circuit and injection of water from non-regular water source) are evaluated in respect of dangerous pressure increase and thermal shock in fuel channels. These results were used during development of severe accident management guidelines for RBMK-1500 at Ignalina NPP.
Piping Supports Design by Analysis and Computer Program Development for NPPs
Authors: Liang Bingbing , Sun Xuewei , Li Gang , Yin Haifeng
Published: July 2009

There are over ten thousand piping supports in a NPP. Piping supports are generally divided into two groups, one is standard and the other is non-standard. No matter to which group the support belongs, should it be proved to withstand the loads acting on it with sufficient design margin following the requirements of ASME or AISC code. With the development of the computer soft ware and hardware, we use 3D mathematic model to do stress analysis for the support. The model is a combination of linear element and nonlinear element, elastic material properties and inelastic material properties. For this composite model, we use plastic limit load analysis to determine assessment section, and use linearizing method to evaluate the section, and use simplified method to model nonlinear boundary conditions. At the end, a computer program is accomplished to perform analysis and evaluation for the supports efficiently and rapidly. The program not only can be used to establish standard support manual, but also can be used to perform analysis and evaluation for the non-standard support in the field.
CFD Assessment of the Local Hot Core Temperature in a Pebble-Bed Type Very High Temperature Reactor
Authors: Kim Min-Hwan , Lim Hong-Sik , Lee Won Jae
Published: July 2009

Assessment of the local hot core temperature during normal operation in a pebble-bed type of Very High Temperature Reactor (VHTR) has been carried out by using the Computational Fluid Dynamic (CFD) method for which the boundary conditions were obtained from the results of a macroscopic analysis of the core using a system thermal analysis code, GAMMA. Three pebble arrangements are selected, which are Simple Cubic (SC), Body-Centered Cubic (BCC), and Face-Centered Cubic (FCC). Results showed that the SC arrangement having the lowest porosity gives the highest fuel temperature of 1237°C but still below the normal operational fuel limit of 1250°C. Comparison of the CFD results with an empirical correlation was made for the pressure drop and the Nusselt number but there were large differences between them. The benchmark calculation of a pressure drop for packed particles in a square channel indicated that the correlation for the full core used in the system code is not appropriate for the prediction of a local thermal fluid behavior.
Application of Magnetic Stir Welding to Dissimilar Metal Structural Weld Overlay
Authors: Yoshida Takeshi , Matsuoka Takaaki , Uchida Yuta , Hirano Takashi
Published: July 2009

Alloy 600 and associated welds, Alloy 82/182 of the Pressurized Water Reactor (PWR) plants have been known as being susceptible to the Primary Water Stress Corrosion Cracking (PWSCC). Dissimilar metal (DM) piping butt welds were welded with Alloy 82/182. As one of the mitigation techniques of the PWSCC, Structural Weld Overlay (SWOL) has been applied to the DM welds, but it has tendency to occur weld cracks on the first layer. One of the reasons of the weld cracks is the sulfur which is highly contained in stainless steel base metals, because old stainless steels would contain higher sulfur (e.g. 0.02%) than later ones. In response to this situation, Magnetic Stir Welding (MSW) was proposed to apply for the first layer of SWOL, and tested to evaluate its weldabilities. MSW has been developed for several years, and it is generally known that MSW has characteristics to improve a heat transfer in the molten pool, so that it could reduce a dilution. The purpose of this study is to evaluate weldabilities of MSW for welding Alloy 52 and/or Alloy 52M as filler metal on high sulfur contained stainless steel pipe. Single bead tests and all position welding tests were conducted. As a result of this study, MSW can prevent from occurring weld cracks and lack of fusion due to stirring effects of the molten pool. Therefore, SWOL can be welded without weld cracks on the first layer by applying MSW, even though the stainless steel base metal contains relatively high sulfur. In addition, MSW can weld at high wire supply rate because of prevention of lack of fusion. So it could improve weld efficiency.
Interfacial Phenomena of Radiation-Induced and Photo-Induced
Authors: Honjo Yoshio , Furuya Masahiro , Takamasa Tomoji , Okamoto Koji
Published: July 2009

When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation induced surface activation (RISA). In order to investigate radiation-induced and photo-induced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to irradiation time before the contact angles reached their super-hydrophilicity state. The reciprocals of contact angles correlate well with integrated intensity by a straight line, regardless of the irradiation intensity and time. Radiation-induced and photo-induced hydrophilicity phenomena are identical to each other in this regard. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly shows that a certain humidity in ambient gas is required to take the place of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the O 1s spectrum, indicating the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface.
Review of Quality Assurance for CMRR at LANL
Authors: Wilde Taunia S. , Holmes Richard A. , Sandquist Gary M.
Published: July 2009

The Chemistry and Metallurgy Research Replacement (CMRR) Project under development at Los Alamos National Laboratory is being executed under a variety of Quality Assurance Requirements, DOE Orders, and CFR Regulations for the design, construction, and testing of all physical aspects of the CMRR Project. This unique US Nuclear Defense Project is the most significant new nuclear weapons infrastructure development since inception of the Manhattan Engineering District Program in the 1940’s. Those unclassified aspects of QA requirements and procedures implemented for this major project are presented and assessed for meeting the demanding requirements imposed to ensure safety, reliability, and success of this vital facility to secure US Nuclear Defense in the 21st century.
Hierarchical Bayesian Model Concerned Uncertainty in Number of Failure Events for Component Failure Rates
Authors: Sato Chikahiro , Yoneyama Mitsuru
Published: July 2009

Components failure rates are important inputs of Probabilistic Risk Assessment (PRA). They are generally estimated from data observed in plant’s experiments. However, number of failure events derived from plant’s experiments might not represent the actual value. This means that true failure rates might have more uncertainty than one estimated from the actual experiments. Therefore it has been necessary to develop statistical Bayesian model which can reflect an uncertainty in data for components failure rates. In this study, the uncertainty in number of failure events is considered as a probability process and the hierarchical Bayesian model is applied to reflect the uncertainty for failure rates. As a result, more appropriate result of PRA is obtained with a state of knowledge about number of failure events for failure rates.
TRU-Fueled VHTRs for Applications Requiring an Extended Operation With Minimized Control and No Refueling
Authors: Tsvetkov Pavel V. , Lewis Tom G. , Alajo Ayodeji B.
Published: July 2009

This paper presents an analysis of TRU-fueled VHTR systems focusing on applications requiring an extended operation with minimized control and no refueling (single-batch mode). As an example of such applications, international deployment opportunities for grid-appropriate VHTR systems could be mentioned addressing demands for electricity, industrial heat and co-generation in those regions where minimized servicing is desirable for various reasons. The study is performed for the hexagonal block core concept within the framework of the ongoing U.S. DOE NERI Project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTRs. The up-to-date analysis has shown reasonable reactivity swings, core life limits with respect to fast fluences and criticality.
Development of BWR Power Uprate Method Based on “Heat Balance Shift” Concept
Authors: Kitou Kazuaki , Chaki Masao , Aoyama Motoo , Yoshikawa Kazuhiro , Sasaki Hiroshi
Published: July 2009

We have developed an innovative power uprate method for boiling water reactors (BWRs) that will increase thermal power by more than 5% without having to replace the high-pressure turbine. Reactor power uprate of nuclear power plants is an efficient plant operating method. Most BWR plants need to replace high-pressure turbines when thermal power is increased to over 5% because the main steam flow rate exceeds the inlet steam flow rate limit of the high-pressure turbine. A conventional power uprate method increases the feedwater and main steam flow rate in proportion to increase in thermal power. We examined a decrease in feedwater temperature instead of an increase in the feedwater and main steam flow rate. Since a decrease in feedwater temperature leads to a smaller main steam flow rate, a power uprate of over 5% can be achieved without replacing the high-pressure turbine. We call this power uprate method the “heat balance shift” method. In the present study, we evaluated the heat balance shift method to determine if it can increase electric power to a level higher than that of the conventional power uprate method without replacing the high-pressure turbine.
Transient Analysis of UO2 and MOX-Fueled Cores Based on Microscopic Reactor Physics
Authors: Takeda Toshikazu , Tagawa Hiroaki , Sano Tadafumi
Published: July 2009

A transient analysis has been performed for UO2 and MOX-fueled light water reactor cores based on Microscopic Reactor Physics, which treats the detailed distribution of temperature and effective cross section within a rod. Conventionally the volume-averaged temperature and the Rowlands’ effective temperature are used to calculate fuel rod-averaged cross sections, and applied to the transient analysis. The present method is considered as a reference and the result is compared with the conventional method for a mini fuel core containing eight fuel rods and a control rod. From numerical results, it is found that the Rowlands’ model underestimates the peak power and the volume averaged model produces rather good peak power results. After 1.0 sec, the Rowlands’ model yields the similar power as the reference, while the volume averaged model yields less power than the reference one.
Development of High Temperature Transport Technologies for Molten Salt and Liquid Cadmium in Pyrometallurgical Reprocessing
Authors: Hijikata Takatoshi , Koyama Tadafumi
Published: July 2009

Pyrometallurgical reprocessing technology is currently being focused in many countries for closing actinide fuel cycle because of its favorable economic potential and an intrinsic proliferation-resistant feature due to the inherent difficulty of extracting weapons-usable plutonium. The feasibility of pyrometallurgical reprocessing has been demonstrated through many laboratory scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high-temperature fluids. In this study, a salt transport test rig and a metal transport test rig were installed in an argon glove box with the aim of developing technologies for transporting molten salt and liquid cadmium at approximately 773 K. It was demonstrated that; using a centrifugal pump, molten salt at 773 K could be transported at a controlled rate from 4 to 8 dm3/min against a 1 m head. The transport behavior of the molten salt was found to be similar to that of water, and could be predicted from their similarity of kinematic viscosity. On the other hand, the transportation of liquid cadmium at approximately 700 K could be controlled at a rate of 0.5 to 1.6 dm3/min against a 1.6 m head using the centrifugal pump.
Visual Observation of Fragmentation Behavior on Molten Material Jet Surface in Coolant
Authors: Uchiyama Yuta , Abe Yutaka , Fujiwara Akiko , Nariai Hideki , Matsuo Eiji , Chitose Keiko , Koyama Kazuya , Itoh Kazuhiro
Published: July 2009

For the safety design of the Fast Breeder Reactor (FBR), it is strongly required that the post accident heat removal (PAHR) is achieved after a postulated core disruptive accident (CDA). In the PAHR, it is important that the molten core material is solidified in sodium coolant which has high boiling point. Thus it is necessary to estimate the jet breakup length which is the distance that the molten core material is solidified in sodium coolant. In the previous studies (Abe et al., 2006), it is observed that the jet is broken up with fragmenting in water coolant by using simulated core material. It is pointed out that the jet breakup behavior is significantly influenced by the fragmentation behavior on the molten material jet surface in the coolant. However, the relation between the jet breakup behavior and fragmentation on the jet surface during a CDA for a FBR is not elucidated in detail yet. The objective of the present study is to elucidate the influence of the internal flow in the jet and fragmentation behavior on the jet breakup behavior. The Fluorinert™ (FC-3283) which is heavier than water and is transparent fluid is used as the simulant material of the core material. It is injected into the water as the coolant. The jet breakup behavior of the Fluorinert™ is observed by high speed camera to obtain the fragmentation behavior on the molten material jet surface in coolant in detail. To be cleared the effect of the internal flow of jet and the surrounding flow structure on the fragmentation behavior, the velocity distribution of internal flow of the jet is measured by PIV (Particle Image Velocimetry) technique with high speed camera. From the obtained images, unstable interfacial wave is confirmed at upstream of the jet surface, and the wave grows along the jet-water surface in the flow direction. The fragments are torn apart at the end of developed wave. By using PIV analysis, the velocity at the center of the jet is fast and it suddenly decreases near the jet surface. This means that the shear force acts on the jet and water surface. From the results of experiment, the correlation between the interfacial behavior of the jet and the generation process of fragments are discussed. In addition, the influence of surface instability of the jet induced by the relative velocity between Fluorinert™ and coolant water on the breakup behavior is also discussed.
PWSCC of Nickel Base Alloys in Vapor Phase Environment of Pressurizer
Authors: Tsuruta Takao , Sato Kenji , Asada Seiji , Kobayashi Takaaki , Okimura Koji , Matsubara Nariyasu
Published: July 2009

PWSCC incidents of Alloy 600 in vapor phase environment of pressurizer have been confirmed at several PWR plants. Vapor phase of pressurizer is filled with vapor from primary water, and the inner surface is covered with liquid film. Chemistry of the liquid film may be different from primary water, and this may cause different PWSCC susceptibility. Therefore the chemistry of liquid film of vapor phase has been investigated using simulated mock-up tests, and PWSCC susceptibility of 152 weld metal and TT600 (SG tube) has been investigated under the chemistry of the liquid film of vapor phase and primary water. According to the result of the chemistry investigation tests using mock-up of pressurizer, the liquid film environment was evaluated as follows: DH2 concentration: 300cc/kg·H2O, B:150ppm, Li<0.1ppb, pH320°C:5.6 under the primary water chemistry condition is DH2 concentration:30cc/kg·H2O, B:1950ppm, Li:3.7ppm, pH340°C:6.9. DH2 concentration of the liquid film is ten times higher and pH is lower than that of primary water. PWSCC susceptibility tests have been performed under the environment of the liquid film and primary water. No PWSCC crack propagation of 152 weld metal is confirmed in vapor phase environment. Crack growth rate of TT600 in vapor phase environment of pressurizer is not particularly high compared with that in primary water environment. It is confirmed that Alloy 690 (152 weld metal) has no PWSCC susceptibility under vapor phase environment of pressurizer. The difference of PWSCC susceptibility for Alloy 600 between vapor phase of pressurizer and primary water environment is not significant.
Application of Outer Surface Irradiated Laser Stress Improvement Process (L-SIP) to Pressurizer as Residual Stress Improvement Method for Alloy 600 PWSCC Mitigation
Authors: Ueda Takeshi , Okimura Koji , Wakabayashi Kazuhiro , Akaba Takashi , Kamo Kazuhiko , Ohta Takahiro
Published: July 2009

Improvement of residual stress is effective in a countermeasure to deal with the stress corrosion cracks in pipe welds. A irradiated laser stress improvement process (L-SIP) will be introduced as a method to improve residual stress inside steel pipes. This work method is to improve inner surface residual stress from tensile stress to compressive stress by irradiating laser beam around the welds of steel pipe and utilizing the temperature differences between inner and outer surface. Recently this method is applied to PWR pressurizer surge nozzle on TRUGA unit 2.
Reducing the Uranium-Plutonium Loading of a Fast Reactor Due to the Use of a Low Neutron Capturing Core Coolant Consisted of Enriched Lead-208
Authors: Khorasanov Georgy L. , Ivanov Anatoly P. , Korobeynikov Valery V. , Blokhin Anatoly I.
Published: July 2009

Long-term scenarios of nuclear energy evolution over the world scale predict deployment of fast reactors (FRs) from 2020–2030 and achievement near 2050 the world installed capacity equal to 1500 GWe with essential increasing the FRs number. For several countries (i.e. Russia, Japan) which policies are based on a sharp increase of nuclear production, at the stage near 2030–2040 when plutonium, Pu, from PWR spent nuclear fuel is consumed, the Pu lack will stimulate minimizing its loading in FRs. The period of Pu deficiency will be prolonged till the years when Breeding Gain (BG) equal to 0.2–0.3 in FBRs is obtained which corresponds to Pu inventory doubling time of 44–24 years. In this paper one of opportunities to minimize fuel loading is considered: it is related with using a low neutron capturing lead isotope 208Pb as a FR coolant. It is known, that natural lead, natPb, contains a stable lead isotope 208Pb having a small cross-section of neutron capture via (n, γ) reaction. In the paper it is shown that the mean cross-section <σ n, γ> of radiation neutron capture by the lead isotope 208Pb averaged upon an Accelerator Driven System (ADS) core neutron spectra is by ∼3.7 – 4.5 times less than the corresponding mean cross-section for a natural mix of lead isotopes natPb. This circumstance allows minimizing the load of a lead fast reactor (LFR) core for achievement its criticality as well as the load of the ADS subcritical core — for achievement its small subcriticality. In using 208Pb instead of natPb in the ADS blanket the multiplication factor of the subcritical core, Keff, could be increased from the initial value Keff = 0.953 up to the value Keff = 0.970. To achieve this higher value of Keff in the same core cooled by natPb an additional amount, about 20–30%, of U-Pu fuel will be needed. The isotope 208Pb content in the mix of lead isotopes, natPb, is high enough, above 52%, and its separation in large amounts (several tons) is expensive but really solvable technical task. In the ISTC #2573 project developed in 2005 with author participation, it is shown that a new laser photochemical technique of lead isotope separation, being developed in future, permits to obtain large quantities of 208Pb under its acceptable price, less than $200/kg.
Blind Prediction of SMART 2008 Seismic Structural Response Test Results
Authors: Varpasuo Pentti , Ka¨hko¨nen Jukka
Published: July 2009

This paper describes the numerical simulation contribution of Fortum Nuclear Services Ltd. to the round-robin blind prediction of SMART 2008 seismic structural response tests to be conducted by Commissariat Energie Atomique in France in spring 2008. In order to assess the seismic tri-dimensional effects (such as torsion) and non-linear response of reinforced concrete buildings, a reduced scaled model (scale of 1/4th) of a nuclear reinforced concrete building is going to be tested in 2008 on AZALEE shaking table at Commissariat a` l’Energie Atomique (CEA Saclay, France). This test, supported by Commissariat a` l’Energie Atomique (CEA) and Electricite´ de France (EDF), will be part of the “SMART-2008” project (Seismic design and best-estimate Methods Assessment for Reinforced concrete buildings subjected to Torsion and non-linear effects). The first part of the project is a blind prediction of the structure behavior under different seismic loadings. It is presented as a contest, opened to teams from the practicing structural engineering as well as the academic and research community, worldwide. This phase will result in the creation of a predictive benchmark, which should allow us to compare and validate approaches used for the dynamic responses evaluation of reinforced concrete structures subjected to earthquake and exhibiting both 3-D and nonlinear behaviors. The objectives of the predictive benchmark are to: 1) Assess different conventional design methods of structural dynamic analyses, including floor response spectra evaluation; 2) Compare best-estimate methods for structural dynamic response and floor response spectra evaluation. In the next analytical phase to be carried out during the year 2009, the prediction contest will be compared to test results at various levels of seismic excitation (including ‘under-design’ and high ‘over-design’ levels), in order to: 1) Quantify variability in the seismic response of the structure and identify contribution coming from uncertainties in input parameters and random variables; 2) Investigate and compare different methods for fragility curves elaboration. The numerical simulation gives the best estimate values for acceleration response spectra values in five specified response points of the model in two perpendicular horizontal directions for base excitation values from 0.05g up to 0.8 g. Also the maximum and minimum values of the stresses and strains in the concrete and in the reinforcement of four vertical walls of the model are to be simulated as well as the acceleration and displacement response time histories at the top of the model for base excitation values from 0.05g up to 0.8 g.
Simulation for Startup Transients in a Natural Circulation Boiling Loop
Authors: Lakshmanan S. P. , Kumar P. Pradeep , Pandey Manmohan , Iyer Kannan N.
Published: July 2009

Startup of natural circulation boiling water reactors is of current interest due to the transients that occur at low pressure and low power conditions. Numerical simulation can be a useful tool for studying startup transients and for devising appropriate startup procedure. In the present work, a numerical model of an experimental test facility has been developed with RELAP5/MOD3.4. This model has been used to devise a startup procedure, which has been successfully implemented in the test facility. Numerical simulations for the startup transients have been done with the RELAP5 model and results have been compared with experimental findings. The nature of oscillations predicted by numerical simulations is similar to that observed in the experiment. This study demonstrated one possible method for developing startup procedures for natural circulation boiling water reactors.
Analysis of Coupled Neutronic-Thermohydraulic Instabilities in Supercritical Water-Cooled Reactor by Lumped Parameter Modeling
Authors: Pandey Manmohan , Kumar M. Ashok
Published: July 2009

The possibility of instabilities in future nuclear reactors cooled by supercritical water is a matter of concern due to sharp changes in thermodynamic properties of coolant within the core. In the present work, a lumped parameter dynamic model of supercritical water-cooled reactor has been developed for analysis of coupled neutronic-thermohydraulic instabilities. The coolant channel is divided into two nodes with a moving boundary between them. The heater wall dynamics is described by a lumped parameter energy balance. Point neutron kinetics with one group of delayed neutrons has been used to model the power dynamics. Simple non-dimensional equations of state have been obtained for evaluating thermodynamic properties. Stability analysis has been done for various values of parameters such as the reactor power, coolant mass flow rate and inlet temperature. Stability maps have been plotted in the parameter planes. Dynamic simulations have been performed in the time domain to study the nature of operating transients. The stability analysis with neutronics is found to be more conservative. Transient simulations without neutronics indicate a supercritical Hopf bifurcation and the existence of a stable limit cycles in the unstable region. However, simulations with coupled neutronics indicate a subcritical Hopf bifurcation and the existence of unstable limit cycles in the stable region. Therefore, the analysis with neutronics is more conservative and shows that the system can be unstable for large perturbations, even if it is stable for small perturbations.
Non-Oxide Ceramics: Chances for Application in Nuclear Hydrogen Production
Authors: Hille Carmen , Lippmann Wolfgang , Herrmann Marion , Hurtado Antonio
Published: July 2009

Research and development are increasingly focusing on the provision and utilization of heat in the high-temperature range above 900 °C, in particular under the aspect of resource-saving energy technologies. On the one hand, the exploitation of the high-temperature range helps to improve the efficiency of energy conversion processes; on the other hand, the provision of high-temperature heat makes it possible to utilize innovative thermochemical processes, which in turn represent environmentally compatible processes. An example to be quoted here is the thermally induced production of hydrogen by the iodine-sulfur process. The high temperatures alone place extremely high requirements on the materials to be used so that metallic materials soon reach their limits of application. If additionally chemically aggressive process media are used, as in the iodine-sulfur process, basically only ceramic materials can be considered as construction materials. In this application, notably silicon carbide (SiC) is favored owing to its excellent high-temperature properties. The possible technical fields of application of such high-performance ceramics can be broadly extended provided that suitable, highly efficient joining methods are available for these ceramics. In addition to its use as a constructional ceramic, SiC can principally also be used as a functional ceramic. For this purpose, the basic ceramic is modified with different additives, providing it with electrical properties that permit its application as a full ceramic heat conductor or sensor. In this case, it also holds true that a suitable joining method for making electrically conductive joints will extend the fields of application considerably. Laser-based joining technologies are being developed for both applications at the Dresden University of Technology. The research work presented here notably focuses on laser joining of electrically conductive SiC ceramics. In addition to a CO2 laser, a diode laser has been used. Basically, electrical connection has been made in two ways. In the first variants, graphite pins are inserted into the joining zone as electrically conductive bridges. In an alternative concept, the oxidic glass filler itself is made electrically conductive with additives. Like that a full ceramic heating conductor joined by means of laser radiation has been tested. The temperature resistance and functionality of the laser-joined heating conductor could be fully demonstrated.
Steam Dryer Acoustic Load Predictions in the Main Steam Line Break Event
Authors: Yan Jin , Bolger Francis , Li Guangjun , Dai Weimin , Klebanov Lev
Published: July 2009

In nuclear reactor design, significant acoustic pressure loads impact the steam dryer hood as a result of the main steam line break outside containment (MSLB) event. When a main steam line breaks, it is assumed that the pipe instantaneously breaks completely open to the ambient environment (double-ended guillotine break). Due to the huge pressure difference between the inside of the reactor pressure vessel (RPV) and surrounding ambient environment, a shock wave will form at the break point and burst into the surrounding environment. At the same time, an expansion wave will travel upstream through the main steam line to the RPV, which results in a pressure reduction on the outside of the steam dryer hood. This expansion wave will create a substantial pressure difference between the two sides of the steam dryer hood with a resultant high stress on the hood. This differential pressure load is the acoustic load used in the structure design evaluations for this event. A key design basis requirement for the steam dryer is to maintain structural integrity during transient, and accident conditions. Demonstration that the steam dryers meet this design basis requires a calculation of the magnitude of the acoustic load on the steam dryer during a MSLB. In this study, Computational Fluid Dynamics (CFD) is used as an alternate calculation method to investigate the phenomenon of MSLB. Transient simulations with fine time steps were carried out. The results show that CFD is a useful tool to provide additional information on the acoustic load as compared to the traditional methods. From the CFD results, the minimum pressure value and its distribution area at different flow times was identified. Through the modeling, an understanding of the detailed transient flow field, particular the acoustic pressure field near the dryer hood during the MSLB was achieved.
A VISION of Advanced Nuclear System Cost Uncertainty
Authors: Taylor J’Tia P. , Shropshire David E. , Jacobson Jacob J.
Published: July 2009

VISION (VerifIable fuel cycle SImulatiON) is the Advanced Fuel Cycle Initiative’s nuclear fuel cycle systems code designed to simulate the U.S. commercial reactor fleet. The code is a dynamic stock and flow model that tracks key material mass flows at the elemental and isotopic levels through the entire nuclear fuel cycle. VISION.ECON is a submodel of VISION that was developed to estimate the costs of electricity. The sub-model uses the mass flows generated by VISION for each of the fuel cycle functions and calculates costs based on the Department of Energy Advanced Fuel Cycle Cost Basis report. This paper provides an evaluation of the cost uncertainty effects attributable to fuel cycle system parameters and scheduling variations. A scenario utilizing a single light-water reactor (LWR) using uranium oxide fuel is examined to ascertain the effects of simple parameter changes. The four variable parameters are burnup, thermal efficiency, capacity factor, and reactor construction time. The effect variables are the total cost of electricity (TCOE) and the fuel cycle costs (FCC). Strategies for future analysis are also discussed. Future work consists of extending the analysis to more complex scenarios, including LWRs using mixed oxide fuel and fast recycling reactors using transuranic fuel.
New Graphical Methodology for Energy Integration in Nuclear Steam Power Plant
Authors: Khoshgoftar Manesh Mohammad Hasan , Amidpour Majid
Published: July 2009

Exergy concept combined with pinch based approach is used for studying the optimal integration of energy conversion systems. The analysis first considers the representation of the hot and cold composite curves of the process and defines the energy and the exergy requirements. Strength of pinch analysis is that system information can be represented using simple diagrams and thus targets for the system under consideration can be readily obtained prior to design. In contrast, the power of exergy analysis is that it can identify the major causes of thermodynamic imperfection of thermal and chemical processes and thus promising modifications can be determined effectively. By combining the strengths of both methods, the proposed method can represent a whole system, including individual units on one diagram, which helps to screen the promising modifications quickly for improving a base case design. This method is Energy Level Analysis. We have developed energy level analysis to energy destruction level as a strategy for energy integration that uses power plant simulation tools to define the interaction between the various subsystems in the plant and a graphical technique to help the engineer interpret the results of the simulation with physical insights that point towards exploring possible integration schemes to increase energy efficiency. In this paper, 1000 MW nuclear steam power plant similar to Bushehr is considered. Simulation of power plant is performed in STEAM PRO software. Computer code is developed to exergy calculation and generation of exergy destruction level representation. In addition, thermoeconomic analysis is performed to generation of other new graphical representation related to exergy destruction that helps us to consider cost rate of destruction in each component.
Global Flow Regime Identification in a Rod Bundle Geometry
Authors: Paranjape Sidharth , Stefanczyk Damian , Liang Yong , Hibiki Takashi , Ishii Mamoru
Published: July 2009

Flow regime maps were obtained for an adiabatic air-water two phase flow through a flow channel with 8×8 rod bundle, which simulated a typical rod bundle in a BWR. Impedance void meters were used to measure the area averaged void fraction at various axial locations in the flow channel The Cumulative Probability Distribution Functions (CPDF) of the signals from the impedance meters were fed to a self organizing neural network to identify the flow regimes. The flow regimes were identified at seven axial locations in the channel in order to understand the development of the flow regimes in axial direction. The experimental flow regime transition boundaries agreed well with the theoretical ones obtained using Mishima and Ishii (1984) model. In addition, the two impedance void meters located across a spacer grid, which were used to study the change in the flow regime across the spacer grid.
Evaluation of Coupling Desalination With PWR Nuclear Power Plant With Pinch, Exergy and Thermoeconomic Analysis
Authors: Khoshgoftar Manesh Mohammad Hasan , Amidpour Majid , Farhadi Ali , Salehi Gholam Reza
Published: July 2009

Exergy concept combined with pinch based approach is used for studying the optimal integration of energy conversion systems. The analysis first considers the representation of the hot and cold composite curves of the process and defines the energy and the exergy requirements. Strength of pinch analysis is that system information can be represented using simple diagrams and thus targets for the system under consideration can be readily obtained prior to design. In contrast, the power of exergy analysis is that it can identify the major causes of thermodynamic imperfection of thermal and chemical processes and thus promising modifications can be determined effectively. By combining the strengths of both methods, the proposed method can represent a whole system, including individual units on one diagram, which helps to screen the promising modifications quickly for improving a base case design. This method is Energy Level Analysis. We have developed energy level analysis to energy destruction level as a strategy for energy integration that uses power plant simulation tools to define the interaction between the various subsystems in the plant and a graphical technique to help the engineer interpret the results of the simulation with physical insights that point towards exploring possible integration schemes to increase energy efficiency. In this paper, 1000 MW PWR nuclear steam power plant is considered. Simulation of power plant is performed in STEAM PRO software. Computer code is developed to exergy calculation and generation of exergy destruction level representation. In addition, thermoeconomic analysis is performed to generation of other new graphical representation related to exergy destruction that helps us to consider cost rate of destruction in each component.
Computational Fluid Dynamics Modeling of Two-Phase Flow Topologies in a Boiling Water Reactor Fuel Assembly
Authors: Tentner Adrian , Lo Simon , Splawski Andrew , Ioilev Andrey , Melnikov Vladimir , Samigulin Maskhud , Ustinenko Vasily , Melnikova Sufia
Published: July 2009

This paper presents recent advances in the development and validation of the two-phase flow topology models implemented in CFD-BWR, an advanced Computational Fluid Dynamics (CFD) computer code that allows the detailed analysis of the two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel assemblies under various operating conditions. The local inter-phase surface topology plays a central role in determining the mass, momentum, and energy exchanges between the liquid and vapor phases and between the two-phase coolant and the fuel pin cladding. The paper describes the topology map used to determine the local inter-phase surface topology and the role of the local topology in determining the inter-phase mass, momentum, and energy transfer. It discusses the relationship between the local interphase surface topology and the traditional channel flow regimes and presents results of experiment analyses in which computed local topologies are aggregated into flow regimes and compared with experimental observations.
Integration of Risk Insights Into Nuclear Power Plant Operations
Authors: Schinzel Glen E.
Published: July 2009

Today’s nuclear plant operator is challenged to safely operate a complex power plant while prudently managing the business aspects with efficiency. Risk insights provide a ready tool to aid today’s operators in effectively performing both of these sometimes contradictory tasks with a sound basis. While plants possess and maintain Probabilistic Risk Assessment (PRA) models, other regulatory applications are readily available to aid the nuclear operator. Some of these tools include 10CFR 50.69 Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, Industry Initiative 4(b) Risk Managed Technical Specifications, and Industry Initiative 5(b) Risk-informed Surveillance Test Intervals. This paper will introduce each of these risk-informed tools and will discuss practical applications of these insights at the South Texas Project nuclear power plant. These insights are readily translatable to other nuclear power facilities. 10CFR 50.69 permits a risk-informed categorization of selected structures, systems, and components. For components determined to be Low Safety Significant, many of the current regulatory controls can be reduced while maintaining reasonable confidence that these ‘Low-ranked’ components continue to perform their design functional requirements. South Texas Project was the industry’s proto-type pilot for this effort. Initiative 4(b) is a risk-informed, configuration-based approach to managing Technical Specification allowed out of service times. The limiting, deterministic allowed outage times are replaced with a Configuration Risk Management Program which uses risk threshold values to determine the length of time a Technical Specification piece of equipment can remain out of service. An imposed back-stop of 30 days is used to limit the allowed outage time. This approach was approved for South Texas Project in July 2007, and South Texas Project was the industry pilot plant for this effort. Initiative 5(b) is a risk-informed approach to Technical Specification surveillance test intervals. This approach allows surveillance test intervals to be removed from Tech Specs and placed in an owner-controlled program. Once relocated, a blending of probabilistic and deterministic insights is used to assess proposed extensions of surveillance test intervals. Once implemented, a feedback process is relied upon to validate the acceptability of the revised testing interval. This approach was piloted by the Limerick Nuclear Station, and South Texas Project submitted a request in October 2007 to the NRC to pursue this initiative. The above risk insights have proven very effective at South Texas Project, and could aid other nuclear operators in making well-founded, informed decisions. Risk insights also allow a Station’s limited resources to be focused on those activities and equipment which are of greatest safety significance. These insights are valuable for current licensees, and may be very beneficial to apply toward new nuclear construction.