bluuebunny/crossref_metadata_2025 huggingface dataset:
Natural Convection With Non-Uniform Heat Generation in Molten Salt Reactor
Authors: Qian Libo , Qiu Suizheng , Zhang Dalin , Su Guanghui
Published: February 2010

The Molten Salt Reactor (MSR) is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The MSR neutronics, such as the distribution of the delay neutron precursors (DNP), is significantly influenced by the fluid flow, which is quite different from the conventional reactors. Therefore, it is very important to do some research on MSR, especially in accident conditions. The present paper studies the natural convection through which the heat generated by the fuel is removed out of the core region (simply a square cavity in this paper). The neutronic theoretical model is founded based on the conservation law, which consists of two-group neutron diffusion equation for the fast and thermal neutron fluxes and that for one-group DNP, in which the convection terms are included to reflect the fuel salt flow. The SIMPLER numerical method was used to calculate the natural convection heat transfer to the molten salt inside a closed cavity for which the boundary temperature was spatially uniform. The equations were discretized by finite volume method based on collocated grids, in which QUICK defect correction was adopted for the convection terms and the central difference was for the diffusion terms. The discretization equations were calculated by ADI (Alternative Direction Implicit) with block-correction technique. The distributions of the dimensionless temperature, the dimensionless velocity, the fluxes and the DNP in the cavity were obtained. The calculated results showed that: a) the distribution of the DNP was correlated both with that of the fluxes and with the fuel salt flow and when Rayleigh number increased, the latter one was of much more importance; b) the distribution of the local Nusselt number varied with different Rayleigh numbers; c) the distribution of the dimensionless velocity and the dimensionless temperature were also closely related to Rayleigh number; d) the maximum dimensionless temperature decreased as Rayleigh number increases.
Heat Exchange Characteristics in Heat Supply and Heat Removal Parts of a Nonisothermal Circuit, Including Temperature and Velocity Fields in a HLMC Flow, for Controlled Impurity Content
Authors: Novozhilova O. O. , Beznosov A. V. , Savinov S. Yu. , Antonenkov M. A.
Published: February 2010

Results of the experimental studies of the heat exchange to the lead heat-transfer agent in the annular clearance in the circulation contour with the controlled and operated processes of mass exchange and mass transfer of the oxygen content are presented. And results of experimental research of lead-bismuth heat-carrier stream velocity structure at a varied content of oxygen content are presented.
Effect of Environmental and Material Conditions on SCC of Carbon Steels in High Temperature Water
Authors: Liu Sue , Wright Michael D.
Published: February 2010

A small number of cracking incidents have occurred in carbon steel outlet feeders of a CANDU® reactor. The cracks were initiated from both the inside and the outside surfaces of tight-radius cold bends. All cracks were intergranular. One plausible explanation for the inside surface initiated cracking is that it is caused by oxidizing water stress corrosion cracking. This mechanism has been studied extensively under boiling water reactor water chemistry (at neutral pH) conditions and the crack path for this kind of SCC is transgranular. It is not known whether the elevated pH in CANDU® primary water chemistry affects cracking behaviour, particularly electrochemical potential thresholds and crack path. Also, the possible effects of cold work have been investigated in the context of SCC of carbon steel in high temperature water. The pronounced cold work in the feeder bends has been identified as the main variable likely to affect the crack path. Based on the understanding of the feeder cracking incidents, an experimental program was carried out to address the effects of coolant pH, and cold work from feeder bending process during fabrication, on the SCC behaviour of carbon steels in the context of an oxidizing water SCC mechanism.
Nuclear Disciplines in Response to Industry and Regulatory Needs Through the European Nuclear Education Network Association
Authors: Safieh Joseph , De Regge Peter P. , Ksuumi Ryoko
Published: February 2010

The European Nuclear Education Network (ENEN), established through the EU Fifth Framework Programme (FP) project, was given a more permanent character by the foundation of the ENEN Association, a nonprofit-making body pursuing an instructive and scientific aim. Its main objective is the preservation and the further development of higher nuclear education and expertise in the nuclear fields. This objective is realized through the cooperation between European universities involved in education and research in nuclear disciplines, nuclear research centers and the nuclear industry. This paper describes the history and structure of the ENEN Association and elaborates on the objectives and activities of its five committees during the first five years of its existence.
Effect of Nano-Scale Steps on the Microscopic Dynamic Contact Angle
Authors: Doi Junpei , Ito Takahiro , Hibi Akira , Kukita Yutaka
Published: February 2010

The motion of the contact line on a solid surface is strongly related with the contact angle there. In these days the relation between them has been revealed with the aid of molecular dynamics techniques; however, most of the studies on such phenomena have been made by assuming the solid surface to be molecularly planer and homogeneous, which condition is not necessarily satisfied in general industries. In this study, we conducted molecular dynamics simulation to investigate the behavior of dynamic contact line with a solid surface accompanied by steps parallel to the contact line. We applied a Couette flow geometry, where two immiscible fluids are confined between two parallel walls, one of which has the steps and is forced to move in the direction normal to the contact line. The behavior of the contact line and the variation in the contact angle were investigated by changing the height of the step from α0 to 4α0 where α0 is the lattice constant of the wall structure. It is shown that in case of the step height identical to α0, the contact line is temporary captured on the edge of the step. On the other hand with higher step the contact line is almost completely pinned on the step edge. The change in the contact angle by the pinning of the contact line on the step edge can be estimated by the macroscopic model as far as the step height is larger than the interaction range between molecules.
Measurement and Analysis for Rewetting Velocity Under Post-BT Conditions During Anticipated Operational Occurrence of BWR
Authors: Sibamoto Yasuteru , Maruyama Yu , Nakamura Hideo
Published: February 2010

A series of experiments was performed for rewetting phenomena on dryed-out fuel rod surfaces under post-BT (Boiling Transition) conditions with high-pressure and high-water flow rate simulating anticipated operational occurrences of a BWR. An analytical model for rewetting velocity, defined by a propagation velocity of a quench front, has been developed on the basis of the experimental results. The rewetting for the post-BT condition is characterized by the faster propagation of the quench front than that for reflood phase conditions during a postulated large-break loss-of-coolant accident. In order to provide an explanation of this characteristic, the present analytical model took an effect of a precursory cooling into account by modifying the existing correlation by Sun-Dix-Tien [1] which is based on a one-dimensional analysis in a flow direction during the reflood phase. The present model demonstrates that the precursory cooling can significantly increase the rewetting velocity by more than an order of magnitude. Applying the experimental correlation developed in the separately conducted experiment into the heat transfer coefficient in the present model at a wet and a dry region with precursory cooling, our data of the rewetting velocity as well as the wall temperature profiles for the variable flow rates are successfully predicted.
The Further Evolution of the EREC-STRESA Web Database
Authors: Blinkov Vladimir N. , Davydov Mikhail V.
Published: February 2010

The EREC-STRESA web-based database contains a large scope of experimental data. It is the unique most complete database resulted from thermohydraulic experiments dedicated to analysis of Russian LWR NPPs safety. This paper presents description of further evolution of the EREC-STRESA Database.
Development and Validation of SACATRI Code for Thermal-Hydraulic Analysis of the CENM TRIGA MARK II
Authors: Merroun Ossama , Al Mers Ahmed , Veloso Marcelo A. , El Bardouni Tarek
Published: February 2010

The Moroccan TRIGA MARK II research reactor operating at power level of 2MW under natural convection regime was acquired by the CENM-Morocco in the aim of developing nuclear technology and its application in the industry and scientific research. Since 2007, start up tests have been performed on this reactor to verify that thermal, mechanical and neutronic parameters fall within the ranges allowing the safe and reliable working conditions of the reactor core. For these purposes, a thermal-hydraulic code named SACATRI, based on sub-channel model, was developed to satisfy the need of numerical simulation tool, able to predict the different thermal-hydraulic parameters of the coolant flow. In the aim of corroborating the physics modeling of the code, the present study is devoted to the quantification of the model accuracy obtained by comparing responses from SACATRI computational model with experimentally measured responses realized on the IPR-R1 TRIGA research reactor.
Two-Phase Cross Flow Between Subchannels in a Tight-Lattice Rod Bundle
Authors: Zhang Weizhong , Yoshida Hiroyuki , Takase Kazuyuki
Published: February 2010

In relation to the thermal-hydraulic design of an innovative Flexible-fuel-cycle Water Reactor (FLWR), this study investigates inter-subchannel cross flow phenomena in a tight-lattice rod bundle. Numerical simulations of cross flow using advance interface tracking method were performed and the results were analyzed by a statistical method to clarify the characteristics of inter-subchannel two phase cross flow in the FLWR reactor core. It was revealed that strong correlation exists between differential pressure and gas/liquid mixing coefficients, and cross flow results mechanistically from differential pressure between subchannels. An approximate model is presented which permits the prediction in detail of the components of the inter-subchannel fluctuation differential pressure. The instantaneous fluctuation of differential pressure between two subchannels in gas-liquid slug flow regime is deemed as a result of the intermittent nature of slug flow in each subchannel. The model is based on the detailed numerical simulation results that pressure drop occurs mainly in liquid slug region and in the bubble region it is negligibly small. The instantaneous fluctuation of differential pressure is associated with pressure gradient in the liquid slug for each channel. In addition to a hydrostatic gradient, acceleration and frictional gradients are taken into account to predict pressure gradient in the liquid slug. This model used in conjunction with the numerical simulation code works satisfactorily to reproduce numerical simulation results for instantaneous fluctuation of differential pressure between two modeled subchannels.
Photographic Study on Bubble Motion in Subcooled Pool Boiling
Authors: Okawa Tomio , Harada Takahiro , Koutsusa Yuta
Published: February 2010

Using a static contact angle as a main experimental parameter, photographic study was carried out to investigate the dynamics of vapor bubbles during subcooled pool boiling on a vertical heated surface. Test fluid was distilled water and the experiments were performed under the atmospheric pressure; liquid subcooling was set to around 5 K. To enable clear observation of bubble behavior with a high speed camera, the experiments were conducted in an isolated bubble regime near the onset of nucleate boiling. Distinctly different bubble behaviors were observed on hydrophobic and hydrophilic surfaces: the bubbles were adhered to the surface for a long period of time when the contact angle was large while lifted off the surface within a short period of time after the nucleation when the contact angle was small. Since buoyancy does not remove the bubble from a vertical surface, mechanisms of lift-off were investigated. It was shown that the change in bubble shape induced by the surface tension force, unsteady growth force and local liquid flow induced by heterogeneous condensation around the bubble are considered to promote the bubble lift-off and the surface tension force acting on the three-phase common line prevents the lift-off. Dependences of bubble size and time at lift-off and the condensation rate on the surface wettability were also investigated.
Next Generation Safety Analysis Methods for SFRs—(2) Experimental Analyses by SIMMER-III for the Integral Verification of the COMPASS Code on Fuel-Pin Disruption and Low-Energy Disrupted Core Motion
Authors: Yamano Hidemasa , Tobita Yoshiharu
Published: February 2010

This paper describes experimental analyses using the SIMMER-III computer code, which is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code. Two topics of key phenomena in core disruptive accidents were presented in this paper: fuel-pin disruption, and low energy disrupted core motion. Related experimental database were reviewed to select appropriate experiments. To analyze the fuel-pin disruption behavior, the CABRI-EFM1 and the CABRI-E7 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall fuel-pin disruption and dispersion behavior, which was characterized by a thermal pin-failure mode, observed in the CABRI-EFM1 experiment. Since the code framework of SIMMER-III cannot treat a mechanical deformation and breakup behavior, the SAS4A code was applied to the CABRI-E7 experiment, where a mechanical pin-failure mode was realized. In this study, such a failure mode was also reasonably simulated. The low-energy disrupted core consists principally of fuel particles and liquid steel (or fuel). Under such a mixture condition, significantly reduced melt penetration length was obtained in the THEFIS out-of-pile experiments. SIMMER-III well simulated the melt freezing and blockage behavior observed in the experiment.
Decontamination of Metal Surfaces Artificially Contaminated With Cs+ Ions by a Laser Ablation
Authors: Won Hui Jun , Baigalmaa Byambatseren , Moon Jei Kwon , Jung Chong Hun , Lee Kune Woo
Published: February 2010

A Q-switched Nd:YAG laser with a 1064 nm and 450 mJ/pulse was employed to study the decontamination characteristics of Type 304 stainless steel specimens artificially contaminated with Cs+ ions. The specimens were treated with KCl and KNO3, respectively. The optimum number of laser shots for the system was determined at a given fluence of 57.3 J/cm2. The relative atomic molar ratio of a metal surface was determined by EPMA. For all the test specimens, more than 95% of the Cs+ ions were removed by the application of 42 laser shots. Cs+ ion removal efficiency could be improved by the addition of nitrate ions to the contaminated metal surface. A specimen treated with a KCl solution was more difficult to decontaminate in the experimental range. Before and after the laser irradiation, the morphology of the metal surfaces was investigated by SEM and XPS. Dusts generated during the laser irradiation were found to accumulate around a crater of the specimen treated with the KCl solution. By analyzing the XPS spectra of the KCl system, it was found that the ablated Cs+ ions formed an oxide in air. The higher decontamination efficiency of the KNO3 system could be attributed to the decomposition of the nitrate ions at a relatively low temperature and the easy reaction of the Cs+ ions with the oxygen generated from the decomposition of nitrate ions.
Corrosion Analysis of Reinforcing Steels in Mortar for Nuclear Facilities
Authors: Nishimura Toshiyasu
Published: February 2010

It is important to estimate the corrosion of reinforcing steel in mortar facilities, because the nuclear plants are located near the sea side. In the case of environmental factors in mortar, the Cl ion concentration and pH were monitored by inserting microelectrodes into artificial pores in the mortar. At the same time, the corrosion behavior of the reinforcing steel was investigated by EIS (electrochemical impedance spectroscopy). In the EIS measurements of the reinforcing steel, diffusion behavior was confirmed in the initial period, but diffusion could no longer be observed after 35 day. In comparison with a 10mm cover thickness, a 20mm cover thickness showed a higher impedance behavior. The Cl ion concentration in the mortar was obtained using Ag/AgCl microelectrodes, showing that this behavior is generally controlled by diffusion. When the diffusion equation was used in this work, the diffusion coefficient (Dc) showed a high value of Dc = 2×10−4 mm2/sec. Similarly, the pH in the mortar was obtained using W/WOx microelectrodes. With a 20mm cover thickness, pH was limited to approximately pH11, but with a 10mm cover thickness, pH continued to decrease to around pH9.5. The latter phenomenon was considered to be the result of neutralization by penetration of the immersion solution from the surface. Based on the results of monitoring with the microelectrodes, solutions simulating those in the pores in mortar were prepared and used in EIS measurements. The charge transfer resistance Rct in the simulated solutions showed good correspondence with the impedance in the low frequency region (2mHz) in the actual mortar. This is attributed to the fact that the corrosion of reinforcing steel was controlled by the solution conditions (mainly Cl concentration and pH) in the pores in mortar. If these solution conditions (Cl concentration, pH) exceed threshold values, it was found that the passivation film is destroyed, resulting in high corrosion.
Study on Thermal Striping at UIS of Advanced Loop Type Fast Reactor: Water Experiment Using a 1/3 Scale 60 Degree Sector Model
Authors: Kobayashi Jun , Kimura Nobuyuki , Tobita Akira , Kamide Hideki , Watanabe Osamu , Ohyama Kazuhiro
Published: February 2010

An advanced loop type sodium cooled fast reactor, JSFR, has been investigated in the frame work of Fast Reactor Cycle Technology Development Project (FaCT). As the temperatures difference between the control rod channels and the core fuel subassemblies is around 100 °C, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). Then, a water experiment was conducted using an 1/3 scale 60 degree sector model. Temperature and its fluctuation intensity distributions around the control rod were measured and an effect of the improved structure against the thermal fatigue was examined.
Results of a Steam-Water Experiment Simulating Two-Phase Flow in Pressurizer Spray Piping
Authors: Oumaya Toru , Nakamura Akira , Takenaka Nobuyuki
Published: February 2010

Pressurizer spray line of PWR plant cools reactor coolant by injecting water into pressurizer. Since the continuous spray flow rate during commercial operation is considered insufficient to fill the pipe completely, there is a concern that a water surface exists in the pipe and may periodically sway. In order to identify the flow regimes in pressurizer spray piping and assess their impact on pipe structure, a two-phase flow experiment under steam-water condition was conducted. In the experiment, temperature fluctuation due to swaying was investigated. With a full-scale mock-up made of stainless steel, two-phase flow of 100°C steam and 60°C water under atmospheric pressure condition was investigated. The experiment revealed continuous temperature fluctuation of 5°C at edge of the water surface. Also, the fluctuation disappeared when concentration of non-condensable gas (Dissolved Oxygen) reached 1ppm. Even though DO concentration reached minimum of 20ppb and the gas phase could be considered almost 100% steam, no violent motion of the surface was observed which was concerned when steam condensation involved. Water levels and values of steady state heat transfer coefficient were estimated and compared with the values by Dittus Boelter correlation.
Next Generation Safety Analysis Methods for SFRs—(7) Potential Model for Classical Molecular Dynamics on Pu-Fe System
Authors: Ito Takahiro , Arima Tatsumi , Himi Masashi
Published: February 2010

Potential model for Modified Embedded Atom Method on Pu-Fe alloy system has been developed to analyze the metal reaction between Pu and Fe with classical molecular dynamics in the framework of the development of the new MPS-based code named ‘COMPASS’ [1] for meso-scale analysis of core disruptive accidents in Sodium-cooled fast reactor. The parameters on Pu-Fe interaction were determined so as to reproduce the properties of PuFe2 intermetallic compound, with applying the potential model for the individual components of Pu and Fe proposed in previous studies. First principle calculations were preformed to compensate for the available experimental data on the material properties of PuFe2. The developed model well reproduces the properties of PuFe2. In the simulation in which the solid Fe is in contact with liquid or solid Pu, formation of crystalline structure of Pu atoms mimicking the Fe structure is observed near the Pu-Fe interface.
Post-Test Investigations of BFBT Critical Power Tests With Trace
Authors: Thieme Marc , Tietsch Wolfgang , Macian Rafael , Sanchez Espinoza Victor Hugo
Published: February 2010

The validation of heat transfer models of safety analysis codes such as TRACE is very important due to the strong interaction of the thermal hydraulics parameters with the core neutronics. TRACE is the reference system code of the US NRC for LWR. It is being developed and extensively validated within the international CAMP-program. In this paper, the validation of heat transfer models of TRACE related to the prediction of the critical power is presented. The validation is based on a large number of critical power tests performed in the NUPEC BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility in Japan. These tests were analysed with the TRACE Version 5 RC 2. The comparison of predictions with the experimental data shows good agreement. The developed TRACE model and the comparison of experimental data with code results will be presented and discussed.
Thermal Fatigue in CANDU Stations
Authors: Yetisir Metin
Published: February 2010

The risk of thermal fatigue, and the resulting pressure boundary failures, have resulted in a significant amount of research in the Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) communities in the past two decades. Thermal stresses arising from thermal stratification phenomena are typically not taken into account in design calculations. Because various forms of thermal fatigue events occurred in safety-related piping, non-design basis thermal fatigue has been the subject of a number of U.S. NRC bulletins that led to a significant amount of research by the nuclear industry. This paper discusses the relevance of known thermal fatigue mechanisms to the CANDU® design (a pressurized heavy water reactor), and thermal fatigue operational experience in CANDU plants. These mechanisms are various forms of thermal stratification phenomena, turbulence penetration and large-scale temperature oscillations in mixing fluid streams.
Measurement of Natural Circulation Two-Phase Flow by Hybrid Doppler Contrapropagating and Pulse Echo Ultrasonic Techniques
Authors: Faccini Jose´ L. H. , Chang Jen-Shih , Harvel Glenn D. , De Sampaio Paulo A. B. , Su Jian
Published: February 2010

In this work, a hybrid ultrasonic technique for diagnostic natural circulation gas-liquid two-phase flow is experimentally investigated. The method is applied on a two-phase natural circulation loop using a contrapropagating transmission flow rate measurement technique in the single phase section and a Doppler velocity measurement and single transducer pulse echo technique in the two-phase region. The two-phase natural circulation loop used in this work consists of a two-phase vertical section where the liquid has an induced flow rate by an air-lift pumping technique. Gas is injected into the lower region of the two-phase test section, and rises to a separation tank causing the liquid to circulate naturally about the loop. The liquid slug velocities, the air flow rates, and the liquid flow rates are measured, respectively, in the two-phase and single-phase section of the loop as well as two-phase interfaces and time averaged void fraction. A comparison with available results from other investigators is given, in terms of liquid superficial velocity and bubble velocity for two flow patterns generated in the loop; bubbly flow and slug flow. The results show that the present hybrid system can be used to identify not only flow regime but also key two-phase flow parameters for natural circulation conditions.
Examination of a Simplified Thermal-Hydraulic Program for a PWR Core Protection System
Authors: In Wang-Kee , Chun Tae-Hyun , Baeg Seung-Yeob
Published: February 2010

A simplified thermal-hydraulic program was developed to quickly calculate the minimum Departure from a Nucleate Boiling Ratio (DNBR) in a pressurized water reactor (PWR) core which is a measure for the core thermal margin. The accuracy of the simplified thermal-hydraulic program would largely depend on the size of the axial nodal sections particularly for the non-iterative scheme for solving the transport coefficient conservation equations. The number of axial nodes in the faster-running DNBR program varies depending on its applications to a PWR core monitoring and protection systems and a reactor safety analysis. It is therefore important to examine the effects of the axial nodal sections in the simplified DNBR program for its application to an advanced PWR core protection system. This paper presents the uncertainties of the minimum DNBR (MDNBR) and the DNBR margin assessment depending on the number of axial nodes. The DNBR margin significantly decreases due to very large MDNBR uncertainty for the number of axial nodes of less than 20.
A Computer-Aided Layout Method for Human Machine Interface of Main Control Room
Authors: Xu Yuqing , Meng Qingxin , Yu Kun , Xu Zhidan , Yan Shengyuan
Published: February 2010

Layout of human machine interface (HMI) plays a very important role in the information transmission between the complex operating and monitoring system and operators. HMI layout of main control room can obviously affect the safety and efficiency of power plant. The unreasonable layout of HMI may influence operators’ judgment and operation under emergency condition, resulting in huge accidents. But up until now, it is difficult to make a all-sided and reasonable layout method for the components of HMI in main control room. For many factors can influence the layout result of HMI, such as importance and usage frequency of each component, which would be changed along with the operation procedure and condition. In this paper, six layout principles of ergonomics are used in the layout process of a computer-aided layout method. Optimizing function is constructed and analytic hierarchy process (AHP) is used in calculation the components weight in layout process. Then, genetic algorithm is used to optimize the layout method. In order to illustrate how the computer-aided layout method is conducted, an example of HMI layout of a power plant is provided.
AP1000: Ansaldo Participation to Engineering and Construction Activities for China Plant
Authors: Frogheri Monica , Saiu Gianfranco
Published: February 2010

Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to the development of AP1000 Plant. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China’s State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. In this framework, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed in the Sanmen site. The contract includes the design of the Steel Containment Vessel, preparation of construction and fabrication Specifications, design and Supply of SCV Mechanical Penetrations, Air Locks and Equipment Hatches. Moreover, Ansaldo Nucleare is in charge of the final design of the AP1000 PRHR-HX and together with Mangiarotti Nuclear will supply the component for the Sanmen Unit 1 NPP. The paper presents an overview of the design and manufacturing activities performed by Ansaldo Nucleare and its partners for the AP1000 plant in China.
Studies of Flow Diversion Due to a Ballooned Pin
Authors: Zeng Yiyun Jessy , Hale Colin P. , Walker Simon P. , Hewitt Geoffrey F.
Published: February 2010

Within the reactor design basis, one of the most significant accidents is a Large Break Loss-of-Coolant Accident (LB-LOCA). In this scenario it is assumed that one of the inlet pipes (“cold legs”), connecting the circulating pump to the reactor vessel, is completely broken so that the pipe ends separate to allow free discharge of the primary coolant from both broken ends. In the case of LB-LOCA inside a pressurised water reactor, four main phases take place, namely blowdown, refill, reflood and long term cooling. During LB-LOCA the interaction between the upward steam-water flow and the ballooned fuel pin is very complex. To better understand the effect of channel geometry on flow diversion a model system has been designed in which air-particle flows are introduced to an annular test section (used to simulate the flow channel of a 3×3 rod bundle). This paper describes particle tracking experiments examining the macroscopic flow and droplet diversions due to a simulated ballooned pin. To determine individual particle trajectories a particle tracking velocimetry (PTV) algorithm has been developed. Results illustrating the 3-D trajectory mapping of individual particles are presented. This provides well-defined experimental data for validation of the models used in the Computational Fluid Dynamics code STAR-CD (CD-adapco) before it is extended to the more complex geometries encountered in fuel rod bundles.
Thermal-Hydraulic Characteristics of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor in Start-Up and Shutdown Processes
Authors: Hu Xunxiang , Dong Yujie , Zheng Yanhua , Zhou Yangping , Shi Lei
Published: February 2010

The start-up and shutdown processes are two important issues of the reactor control and operation for high temperature gas-cooled reactor plants. The analysis of thermal-hydraulic characteristics in these two processes can provide useful reference to the reactor design and manufacture, especially to the final establishment of the operation procedure. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the start-up and shutdown processes are studied by the help of two computer codes THERMIX and BLAST. The key parameters of these two processes, including the reactor power, the core inlet and outlet helium temperatures, the primary pressure and flow rate, as well as the secondary water flow rate and the outlet steam temperature, are analyzed in detail. Moreover, some uncertainties, such as the adjustment of power and secondary water temperature in the start-up process, the flow rate adjustment between the primary helium and the secondary water in the shutdown process, etc., are analyzed to evaluate the influences on the key parameters. The analysis results indicate that the pre-established start-up and shutdown rules are reasonable.
Performance Evaluation of Japanese Nuclear Power Plant Based on Open Data and Information
Authors: Nagata Tadahisa , Sugiyama Ken-ichiro
Published: February 2010

The operation term of Japanese nuclear power plants is less than 13 months. Moreover, the refuel outage period is longer than the other countries. It is likely that excessive preventive maintenances result in early “infant mortality” failures (early failures). However, the statistical evaluation report about failure type of Japanese nuclear power plants was not found. Therefore, Evaluation of plant engineering performance was tried on open data/information with a statistical method. The Weibull distribution/analysis, which needs not only failure data but also censored data like the preventive maintenance, was applied to plant performance evaluation. As open data/information, an annual report; “Operational Status of Nuclear Facilities in Japan” is very popular for the operation database of Japanese nuclear power plants. However, maintenance date/information of failed equipment which caused plant shutdowns were not reported in this annual report. Therefore, every equipment was assumed to be maintained during every shutdown, because this assumption generally makes conservative results of failure rate and Mean Time Before Failure (MTBF) for discussion about early failures. Data till March 2007 were collected from these annual reports. As a result of plant performance evaluation, failure type was early “infant mortality” failure type. Excessive maintenances probably resulted in early failure type. Moreover, influence of early failures and effects of plant design modifications were evaluated. Almost 30% of failures occurred within one month after restart, and failure type excluded these failures was chance (random) type. Early failures within one month may be closely evaluated to improve maintenances. Design modification effect of Japanese nuclear power plants was confirmed.
Using a Real Coded Genetic Algorithm to Obtain the Optimal Parameters of a Cascade of Gas Centrifuge
Authors: Minuchehr Hamid , Zolfaghari Ahmad , Makarachi Peymaan , Noroozy Ali
Published: February 2010

A minimum total number of centrifuges can be taken as an optimization criterion in designing a cascade of centrifuges. The separable power of a centrifuge of given geometry (rotor height and radius) and specified peripheral speed (the highest that materials of construction can withstand) depends upon the internal variables controlling the component drives and upon the operating variables, which are the cut θ, and the throughput, L. To enhance separative power of a gas centrifuges cascade, one needs to optimize all parameters. In this paper, the Real Coded Genetic Algorithm, RCGA, is implemented. As an example of the method, the dependence of separation factor, α to variables cut and throughput is assumed as a typical function. The outcomes of the method are in good agreement with published data.
PWR Fuel Management Optimization Using a New Integer Coded Genetic Algorithm
Authors: Zolfaghari Ahmad , Minuchehr Hamid , Noroozy Ali , Makarachi Peymaan
Published: February 2010

The objective of this paper is to develop a new genetic algorithm (GA) for designing the loading pattern (LP) for pressurized water reactors (PWR). Because of huge number of possible combinations for the fuel assemblies (FA’s) loading in a core, finding the optimum solution is truly a complex problem. In common genetic algorithm the mutation and crossover techniques are used to optimize an objective function but in this paper a new modified crossover along a unique technique is presented. In this study flattening of power inside a reactor core is chosen as an objective function. To obtain optimal FA arrangement, a core reload package code, MAKGA, is developed. This code is applicable for all types of PWR core having different geometries and designs with an unlimited number of FA types. The result is well improved in comparison with pattern proposed by designer.
Ti3SiC2 as a Candidate Material for Lead Cooled Fast Reactor
Authors: Utili M. , Agostini M. , Coccoluto G. , Lorenzini E.
Published: February 2010

One of the main issues regarding the design of LFR is the corrosion-erosion behaviour of materials in contact with high temperature and high velocity lead, such as pump impeller and bearing materials. Those materials will worked in Lead at about 480°C with a velocity relative to impeller blade in the magnitude of 10m/s. Ti3SiC2, a ternary compound was selected as one of the most promising candidate material, the most significant aspect of Ti3SiC2 is that it combines some of the most attractive proprieties of ceramics with those of metals. This material was already applied in industry but more effort is needed to qualify its performance when compared with candidate structural stainless steel available in Europe. With the purpose of evaluating the corrosion resistance of Ti3SiC2, a preliminary screening test was carried out at ENEA Brasimone in fluent lead under oxidation conditions (oxygen concentration of about 10−6 wt% in the melt), with an average temperature of 500°C and a stream velocity of about 1 m/s.
Thermal Aspects for Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors
Authors: Villamere Bryan , Allison Leyland J. , Grande Lisa , Mikhael Sally , Rodriguez-Prado Adrianexy , Pioro Igor
Published: February 2010

SuperCritical Water-cooled Reactors (SCWRs) are a Generation IV nuclear reactor concept. Two main SCWR design concepts are Pressure-Vessel (PV) type and Pressure-Tube (PT) type reactors. SCWRs would use light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). A reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is that a SCW NPP will have a thermal efficiency of 45 to 50%, a remarkable improvement from the current 30–35%. SCWRs have another added benefits such as a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing an SCW CANDU reactor. This concept refers to a 1200-MWel horizontal pressure-tube type reactor with the following operating parameters: a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Materials and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for a fuel is an enriched Uranium Dioxide (UO2). The industry accepted limit for fuel centreline temperature is 1850°C, and previous studies have shown that the fuel centreline temperature of UO2 pellet might exceed this value at certain conditions. Therefore, a thermal conductivity of the fuel must be sufficiently high to transfer large heat flux within a fuel pellet. Also, a sheath material must withstand supercritical pressures and temperatures inside aggressive medium such as supercritical water, so it should be corrosion-resistant, high-temperature and high-yield strength alloy. In general, sheath materials in various SCWR concepts have a temperature design limit up to 850°C. Uranium Carbide and Uranium Dicarbide are excellent fuel choices as they both have higher thermal conductivities compared to conventional nuclear fuels such as uranium oxide, MOX and Thoria. UC and UC2 are high-temperature ceramics. The sheath material being considered is Inconel 600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. To model a generic SCWR fuel channel, a 43-element bundle string was used. In this paper, bulk-fluid, sheath and fuel centreline temperature profiles together with heat transfer coefficient (HTC) profile were calculated along the heated length of a fuel channel. Also, selected thermophysical properties of various nuclear fuels are listed in the present paper.
Study of the Influence of Surface Finishing on the Corrosion/Oxidation Behaviour of Stainless Steels in Stagnant Lead Bismuth Eutectic
Authors: Marti´n-Mun˜oz F. J. , Soler-Crespo L. , Go´mez-Bricen˜o D.
Published: February 2010

Lead-bismuth eutectic (LBE) is of interest as a coolant in the design of fast reactors and also as both a coolant and a spallation target in proposed transmutation schemes for radioactive waste. However, liquid metal corrosion to the structural materials can proceed via various processes: species dissolution, formation of the inter-metallic compounds at the steels/liquid metal interface... It is known that the formation of an oxide scale on a steel surface can protect it dissolution attack by liquid LBE. The variables that play an important role on the feasibility of the formation of an oxide layer and on its protective characteristics for steels in contact with LBE are oxygen content, temperature, materials composition and evolution with time, but also surface state of steels prior to testing or weld joints, being these parameters not very widely studied. For the study of the influence of the surface finishing, specimens with different surface states were prepared (as-received, grinded, grinded and polished and electrolitically polished). These treatments gave to the materials a different degree of cold working, the higher for the mechanised samples and the lowest for the electrolitically polished. Tests were carried out at 500 and 550°C from 100 to 2000 hours with two different H2/H2O ratios: 3 and 0.03. The general conclusion is that the effect of surface finishing on the corrosion/protection processes is not significant for the conditions tested.
Dependence of Slow Crack Growth Resistance on Pipe Sizes for Scratched Polyethylene Nuclear Water Pipe
Authors: Zhou Z. Jimmy , Chang Dane
Published: February 2010

Polyethylene (PE) material is chemically inert and does not corrode. It is a preferred material choice for many piping applications including the service water pipe in nuclear power plants. With an outstanding field service record for 5 decades for both gas and water distribution, recently a 36″ SDR 9 PE4710 pipe received approval for Release Request from the U.S. Nuclear Regulatory Committee (NRC) and was installed in 2008 by a major US utility to replace the steel pipe for the safety related nuclear water pipe application. The PE pipe industry uses a maximum 10% scratch as a rule of thumb for scratched pipes. As the pipe diameter increases, the stress intensity increases for the same 10% of scratch depth. It is a concern for the regulators how the increased stress intensity affects the long term performance of the scratched pipe against slow crack growth (SCG) failure. The SCG resistance depends on the resin material, stress intensity and testing temperature. The stress intensity is controlled by pipe geometric factors, the applied stress, and the notch depth. In order to achieve the equivalent SCG performance for all pipe sizes, the following actions may be taken: (1) use a resin material that has higher SCG resistance to compensate impact from the increased stress intensity; (2) reduce the scratch depth and (3) reduce the pipe hoop stress so that the increased stress intensity is decreased. The minimum SCG resistance requirement that eliminates the need for action (2) and (3) needs to be determined. ASTM F1473 has addressed the equivalent pipe sizes from 1″ SDR 11 to 8″ SDR 11 pipe. This paper investigates the dependence of SCG resistance on the geometric factors for the pipe diameters from 2″ to 44″. The recent PE4710 nuclear water pipe installation is also discussed.
Long-Term Strength and Design of Polyethylene Compounds for Pressure Pipe Applications
Authors: Boros Stephen J.
Published: February 2010

The interest in using polyethylene pipe in Class 3 safety water systems in nuclear power plants has grown tremendously in the last few years. PE pipe brings a host of benefits to the application in the form of long-term performance and reliability due to not being prone to corrosion and tuberculation. As the work continues through various ASME committees to develop the appropriate code language for the design and use of PE pipe, it is clear that plastics are not evaluated the same way metallic components would be in similar applications. However, the nature of the failure (i.e. ductile or brittle) is important for both. This paper will give an overview of the methodology used to establish the long-term hydrostatic strength of polyethylene compounds, and how that strength is used for engineering design in a safe a reliable manner. The strength of a polyethylene compound, being a thermoplastic, cannot be determined from a short-term tensile strength test, as with most metals. As such, testing and evaluation methodologies have been developed which take into account the viscoelastic creep response of thermoplastics, as well as potential changes in failure mode, in order to forecast the long-term hydrostatic strength of these materials so they can be safely used in a pressure pipe application. Since PE was first used in a piping application in the late 1950s, PE has continued to evolve as have the methodologies used to evaluate its strength against stresses induced by hydrostatic pressure. The common method for evaluation relies on putting specimens under multiple continuous, steady-state stress levels until failure. These data points are then used in a log-log linear regression evaluation. This regression equation is then extrapolated to a point sufficiently further out in time to where a long-term strength can be established. It has been clearly established that over a temperature range that the stress rupture behavior of PE follows an Arrhenius, or rate process, relationship between temperature and strength. By testing at elevated temperatures it can be “validated” that the extrapolation remains linear and ductile beyond the actual test data. This and other criteria established by ASTM D 2837 and the Plastics Pipe Institute’s Hydrostatic Stress Board allow for establishing an appropriate maximum working stress that will assure a very long design life.
Verifications of Diffraction Modeling Capability in a Coastal Spectral Wave Model
Authors: Zheng Jinhai , Tang Yu
Published: February 2010

WABED (Wave Action Balance Equation with Diffraction) is a 2-D coastal spectral wave process model and used in the development of a practice-oriented random wave prediction for coastal engineering studies at inlets, navigation projects, and wave-structure interactions. Wave diffraction is implemented by adding a term derived from the parabolic wave equation to the wave action balance equation. This paper describes the evaluation of the modeling capability to represent wave diffraction available in the WABED, which is accomplished by testing the present model for diffraction-prone cases such as the wave field of a gap in an infinitely long breakwater and that over an elliptical shoal. Computations are compared with experimental observations and Sommerfeld’s analytical solutions. Comparisons indicate that WABED performs reasonable well in these conditions and is capable of predicting wave diffraction effectively together with refraction and shoaling.