bluuebunny/crossref_metadata_2025 huggingface dataset:
The Implementation of the Dynamic-Order-Extended Time-Delay Dynamic Neural Units to Heat Transfer System Modelling
Authors: Simeunovic´ Goran , Bukovsky Ivo
Published: July 2009

The paper introduces linear dynamic-order extended time-delay dynamic neural unit (DOE TmD-DNU) whose adaptation by the dynamic backpropagation learning rule is enhanced by the genetic algorithm. DOE TmD-DNU is a possible customization of novel class of artificial neurons called time-delay dynamic neural units (TmD-DNU). In standalone implementations, these artificial dynamic neural architectures can be viewed as analogies to continuous time-delay differential equations, where the equation parameters are unknown and are adaptable such as neural weights and other parameters of artificial neurons. Time delays on neural inputs of a unit and in the state feedback of a unit are also considered as unit’s adaptable neural parameters. These new neural units equipped with adaptable time delays can identify all parameters of a continuous time-delay dynamic system including unknown time delays both in the unit’s inputs as well as in its state variables. Incorporation of adaptable time delays into neural units significantly increases approximation capability of individual neural units. It results in simplification of a neural architecture and minimization of the number of neural parameters, and thus possibly in better understanding the obtained neural model. It has been shown, that stable adaptation of all parameters of TmD-DNU including time delays can be achieved by dynamic modification of backpropagation learning algorithm. However, sometimes the relatively slow convergence rate of the neural parameters and the convergence rather toward local minima of error function can be considered as drawbacks of the adaptation. This paper focuses the improvement of the backpropagation learning algorithm of TmD-DNU by the genetic algorithm and its application to heat transfer system modeling. The adaptation learning algorithm based on the simultaneous combination of dynamic backpropagation and genetic algorithm has been designed to accelerate the convergence of time-delay parameters of a neural unit and to achieve the global character of minimization of error function. The neural weights and parameters, except the time-delays, are adapted by dynamic modification of backpropagation learning algorithm, and those that represent time-delays can be adapted by the genetic algorithm. Results on system identification of an unknown system with dynamics of higher-order including unknown time delays are shown in comparison to achievements by common identification methods applied to the same system. The robust identification capabilities, the aspects of network implementation of TmD-DNU, and the prospects of their nonlinear versions, i.e. higher-order nonlinear time delay dynamic neural units (TmD-HONNU) are briefly discussed with respect to the learning technique presented in this paper.
New Metallic Gaskets Qualification: Resistance to Accident and Normal Conditions in TN International Casks
Authors: Marlier Re´gis , Issard Herve´
Published: July 2009

For dual purpose “transport/storage” casks (like the TN24 casks family or casks for vitrified wastes), the confinement by way of metallic gaskets has long been the chosen solution. Their metallic nature leading to a good behaviour in long-term and high temperature conditions, they still have to satisfy the transport regulations requirements, which are not a given. In order to secure the supplying of such strategic parts of a cask and, above all, to keep developing efficient sealing solutions, TN International has found and qualified new metallic gaskets. For this qualification, a series of tests complying with the accident and normal conditions of transport/storage have been designed and performed on these new metallic gaskets. Indeed, the gaskets have to withstand three major types of “accidental” conditions: 1/ parallel sliding of the lid and the body, resulting in a lateral displacement of the contact surfaces, 2/ separation of the lid and the body, resulting in a decompression of the gasket, 3/ 30 minutes-800°C fire on the cask, resulting in a peak temperature of the gasket. They also have to withstand normal transport/storage conditions, that is to say thermal ageing for long period. Finally we wanted to know how these seals resisted to small particles present during the compression. Tests assemblies and procedures have been developed, in accordance with TN International cask design. Then lateral quasi-static displacement tests up to 3-mm, lateral dynamic displacement tests up to 3-mm, multiple decompression and recompression tests, over-400°C peak temperature tests, long term ageing tests and particle-resistance tests were carried out. The dimensional, visual and leaktightness results of these conservative and highly severe tests were fully satisfactory and allow to establish the safety and efficiency of these new metallic gaskets.
Thermo-Mechanical Analysis of Residual Stresses in Arc Welded Thin-Walled Cylinders
Authors: Qureshi Ejaz M. , Malik Afzaal M. , Dar Naeem Ullah
Published: July 2009

The problem of reduced strength of the structures in and around the weld zone due to the residual stresses in arc welding process is a major concern of the welding industry for decades. The prediction of the transient and residual stress fields during and after the completion of the welding process is of critical importance to ensure the structural integrity of welded structures. Circumferentially welded thin-walled cylinders are widely utilized in many engineering applications including the oil transportation system, boiler and pressure vessel industries. This paper presents a detailed three-dimensional finite element (FE) study to investigate the welding induced residual stresses in circumferentially welded thin-walled cylinders. The complex phenomenon of arc welding is numerically solved by sequentially coupled transient, non-linear thermo-mechanical analysis approach to simulate Gas Tungsten Arc (GTA) Welding process. Single pass butt-welded geometry with single “V” groove for two 300 mm outer diameter cylinders with 3 mm wall thickness and 150 mm length are used in numerical simulations. Temperature dependent thermo-mechanical behavior for low carbon steel is modeled and filler metal deposition by using the element birth and kill features is incorporated. Widely accepted double ellipsoidal GOLDAK heat source model for arc welding is introduced and implemented. Simulation of the entire welding process is accomplished by using author written subroutines in ANSYS®, general purpose FE software. The accuracy of the developed FE simulation strategy is validated with experimentation for temperature distribution and residual stresses.
Energy Spectral Properties Due to Bubble-Liquid Interaction in Oscillating-Grid Turbulence
Authors: Morikawa Koichi , Urano Shigeyuki , Sanada Toshiyuki , Saito Takayuki
Published: July 2009

In the present study, liquid-phase turbulence modulation induced by bubble swarms ascending in arbitrary turbulence is experimentally investigated. Liquid-phase homogeneous isotropic turbulence is formed using an oscillating grid in a cylindrical acrylic vessel of 149 mm in inner diameter. A bubble swarm of 19 bubbles is examined. The bubble size and generating time are completely controlled applying a bubble generating device using audio speakers. The bubble swarms are generated at a frequency of 4 Hz. The motion of individual bubbles of the bubble swarm is captured by a high-speed video camera. The liquid phase motion is also measured by PIV. The turbulence intensity, spatial correlation, integral scale and energy spectrum are calculated from the PIV data. It is confirmed that the oscillating-grid turbulence is an isotropic turbulence from the PIV result. When the bubble swarm is added, the original isotropic turbulence is modulated to an anisotropic turbulence by the mutual interaction between the bubble swarm and ambient isotropic turbulence. Moreover the bubble motion is notably changed. The bubble swarm contributes an increase of the energy spectrum to the ambient oscillating-grid turbulence. Moreover the suppression of energy spectrum occurred by interaction between bubble swarm and oscillating-grid turbulence compared with the energy spectrum in the flow induced by bubble swarms.
Development of Code Case N-755: Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities
Authors: Schaaf Frank J.
Published: July 2009

With the increasing failures of metallic pipe in nuclear Service Water Systems, a new material needed to be found. One option is polyethylene (PE) pipe. PE pipe can be used in non-safety applications at a nuclear plant using the American Society of Mechanical Engineers (ASME) B31, Standards of Pressure Piping with no regulatory review. However, the use of PE material in safety applications, which are regulated by the Nuclear Regulatory Commission (NRC), necessitates a new Standard with special requirements. At the request of the Duke Power Corporation, a new ASME Standard was written by a special Project Team. This standard is found in the form of a Code Case under the control of the ASME Boiler & Pressure Vessel Code (B&PVC). The Code Case utilizes Sections of the B&PVC as its foundation and includes the design, procurement, installation, fusing, examination and testing requirements for the use of PE pipe within safety systems. The first version of the Code Case contained only the minimum requirements needed to support Duke Power Corporation’s first phase of PE piping installation into a safety system within a nuclear power plant. The Code Case developed is titled, N-755, Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities. The first version of this case is limited to buried piping using only the following components; straight PE pipe, PE mitered elbows, and transition flanges. The Code Case will be revised as data for material and components becomes available at the completion of testing.
Future Fuel Cycles: A Global Perspective
Authors: Duffey Romney B.
Published: July 2009

Nuclear energy must be made available, freely and readily, to help meet world energy needs. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world’s largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Canada’s Atomic Energy of Canada Limited (AECL) has developed a unique world-class nuclear power reactor technology — the CANDU® reactor based on the Pressure Tube Reactor (PTR) concept, moderated by heavy water (D2O), also sometimes called the Pressurized Heavy Water reactor or PHWR. With expectations of significant expansion in nuclear power programs worldwide and the resultant concerns about uranium availability and price, there is a growing desire to improve resource utilization by extracting more energy from each tonne of mined fissionable material. Attention is therefore being increasingly focused on fuel cycles that are more energy efficient, reduce waste streams and ensure sustainable futures. There are also many compelling reasons to utilize advanced fuel cycles in PTR (CANDU-type) thermal spectrum reactors. Because of its inherent technical characteristics, PTRs have a great deal of fuel cycle flexibility. The combination of relatively high neutron efficiency (provided by heavy water moderation and careful selection of core materials), on-line fuelling capability and simple fuel bundle design mean that PTR reactors can use not only natural and enriched uranium, but also a wide variety of other fuels including thorium-based fuels and those resulting from the recycle of irradiated fuel. In addition, the PTR can be optimized as a very effective “intermediate burner” to provide efficient fuel cycles that remove residual minor actinides. This inherent fuel cycle flexibility offers many technical, resource and sustainability, and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and intent is to be consistent with improved or enhanced safety, licensing and operating limits and global proliferation concerns, and sustainable energy futures.
Numerical Simulation of Radiolysis Gas Detonation in BWR Exhaust Pipes and Mechanical Response of the Piping to the Detonation Pressure Loads
Authors: Kuznetsov M. , Lelyakin A. , Breitung W.
Published: July 2009

1-D numerical calculations of radiolysis gas detonation and mechanical response of a 12.5-m long BWR exhaust pipe have been performed. To reproduce one of scenarios of steam condensation with following radiolysis gas accumulation in an exhaust pipe initially filled with nitrogen at an initial pressure of 1.6 bar and temperature of 35°C, nitrogen diluted radiolysis gas mixtures were used for the numerical simulations. Nitrogen concentration in radiolysis gas composition was changed in the range of 0–80 mol.%. It permits to significantly change detonation properties of the test mixtures. Different gas dynamic effects such as a precursor shock wave and shock reflection on the maximum detonation pressure have shown in our calculations. Maximum pressure load of the piping can be achieved near the deflagration-to-detonation transition (DDT) point and at the tube end. At those positions the maximum pressure could be at least 2–2.5 times higher than the CJ detonation pressure. Dilution of the radiolysis gas mixture with nitrogen leads to reduction of the radiolysis gas detonability and to increase of run up distance to the DDT point. In this case so called “late detonation initiation” in a pre-compressed zone close to the reflection end can occur. It produces extremely high pressure load which can locally be 10 times higher than CJ-detonation pressure for steady state detonation. Mechanical response of the 12.5-m long austenitic steel pipe (Werkstoff Nr. 1.4541) with a diameter 510 mm and wall thickness of 15 mm was calculated for different detonation pressure loads. Maximum deformations of the pipe were obtained close to the DDT point and at the reflection end. It was shown that even for the worst case mixture with a “late detonation initiation” the deformation of tested pipe is very low (not more than 0.2%) and no danger exists for the integrity of the exhaust pipe under radiolysis gas detonation load. This 1D numerical code permits for the first time a continuously mechanistic analysis of the complicated processes with DDT and detonation propagation in closed pipes and it can be used for designers and for piping safety analysis.
Numerical Simulation of Fluid Flow and Heat Transfer in the Advanced CANDU® Reactor Endshield Using ANSYS-CFX and Porous Media Approach
Authors: Khair Khairy , Baset Saleh , Dimitrov Livia , Millard Julian
Published: July 2009

A distinguishing design feature of CANDU® nuclear reactors is the use of horizontal fuel channels housed in a horizontal vessel called the calandria vessel, which is made of stainless steel. The calandria vessel has two endshields (each consists of two tubesheets called calandria tubesheet and fuelling tubesheet), which provide supports for the fuel channels among other purposes. The two tubesheets of each endshield are joined by a series of stainless steels tubes called lattice tubes. The space within each endshield between the tubesheets and the outside of lattice tubes is filled with cooling water and carbon steel balls. Thus, the endshields provide shielding to reduce radiation reaching the fuelling machine vaults. Nuclear heat is generated within the endshields. Endshields also receive heat from the primary heat transport system by conduction through the fuel channel bearings, by conduction and radiation through the annular insulating gap for the lattice tubes, and by convection from feeder cabinet. Three finite volume models have been developed to simulate different aspects of the coupling between the fluid flow and thermal energy. In model 1, the whole space inside the endshield is modeled as double porous medium to represent the lattice tubes and the steel balls regions respectively. In model 2, the lattice tubes are modeled in details and a single porosity is used to model the space occupied by the steel balls only. This detailed model also predicts the temperature on the surface of the lattice tubes. The work presented in the paper shows that the results from both models are in good agreement. It also shows that the current design of the ACR® endshield cooling satisfies the design requirements with respect to the heat transfer to the shield cooling system during normal operation. Model 3 is used to predict temperature and flow behaviour under transient load service conditions.
Oriented Spray Cooling System Ultimate Heat Sink for Future Nuclear Plants
Authors: Bowman Charles F.
Published: July 2009

The reference safety-related ultimate heat sink (UHS) for the evolutionary advanced light water reactor nuclear plants that require a safety-related reactor service water system (RSWS) is a spray pond. Spray ponds offer significant advantages over mechanical draft cooling towers including superior simplicity and operability, lower preferred power requirements, and lower capital and maintenance costs. The UHS for the Columbia Generating Station (CGS), one of the last nuclear plants that was licensed during the last round of nuclear plant construction in the United States of America (USA), is the Oriented Spray Cooling System (OSCS), an evolutionary spray pond design. Unlike a conventional spray pond in which spray nozzles are arranged in a flat bed and spray upward, the OSCS nozzles are mounted on spray trees arranged in a circle and are tilted at an angle oriented towards the center of the circle. As a result, the water droplets drag air into the spray region while the warm air that is concentrated in the center of the circle rises. Both of these effects work together to increase air flow through the spray region. Increased air flow reduces the local wet-bulb temperature (LWBT) of the air in the spray pattern, promoting heat transfer and more efficient cooling. During the late 1970’s while working for the Tennessee Valley Authority, the author developed a purely analytical model to predict the thermal performance of the OSCS which was successfully compared with the OSCS at the CGS in the mid 1980’s. This paper describes the OSCS and the analytical model that is used to predict its performance and compares the predicted performance of the OSCS at the CGS with the full-scale field test results. The paper describes how this technology has been successfully used to design the UHS for a future nuclear plant that requires a safety-related RSWS that must not exceed a peak temperature of 95 °F.
Effect of Orientation on Mechanical Behavior of an Extruded Al Alloy
Authors: Malik M. A. , Salam I. , Muhammad W.
Published: July 2009

The extruded materials are extensively used in chemical, food and nuclear industry and generally offer a unique combination of strength and freedom with regard to design solutions. During extrusion, material flow occurs in the direction of applied force and as a result microstructure change. The process ultimately induces variation in the mechanical properties when tested along or across the extrusion direction. The uniaxial tensile test is a simple and versatile test to expose most of the mechanical properties of the materials required to ensure the reliability of the systems. In present study, the mechanical behavior of an Al-Mg-Si alloy extruded cylinder has been determined with the help of uniaxial tensile test in longitudinal and transverse orientations. The microstructural features revealed significant difference in two orientations and constituent particles were found aligned in the direction of extrusion. Tensile tests were conducted in displacement mode at different cross head speeds corresponding to strain rates ranging from 10−5 to 10−1 s−1. The tests were conducted at ambient temperature in air atmosphere. The data thus obtained include: yield strength, ultimate tensile strength, percent elongation and reduction of area. Comparing the trends of strength variation, the material shows higher yield strength in longitudinal orientation as compared to transverse orientation. A slight increase in the yield strength with increasing strain rate was found in both the orientations. The ultimate tensile strength in both the directions was found similar and there was no appreciable change with increasing strain rate. The elongation and reduction in area were found higher in the longitudinal orientation. The effect of strain rate on these properties was negligible up to maximum speed tested. In longitudinal orientation typical dimpled fracture was observed indicating deformation before failure. In transverse orientation shallow dimples were present. The present study revealed that the distribution of constituent particles in an extruded thick-walled cylinder has a pronounced effect on its mechanical behavior and fracture morphology.
Effects of Surface Tension on Two-Phase Void Drift Between Triangle Tight Lattice Subchannels
Authors: Kawahara Akimaro , Sadatomi Michio , Higuchi Tatsuya
Published: July 2009

In this study, void drift phenomena, which are one of three components of the inter-subchannel fluid transfer, have been investigated experimentally and analytically. In the experiments, data on flow and void redistributions were obtained for hydraulically non-equilibrium flows in a multiple channel consisting of two subchannels simplifying a triangle tight lattice rod bundle. In order to know the effects of the reduced surface tension on the void drift, water and water with a surfactant were used as the test liquids. In addition, data on the void diffusion coefficient, D˜, needed in a void drift model, have been obtained from the redistribution data. In the analysis, the flow and the void redistributions were predicted by a subchannel analysis code based on a one-dimensional two-fluid model. From a comparison between the experiment and the code prediction, the present analysis code was found to be valid against the present data if newly developed constitutive equations of wall and interfacial friction were incorporated in the model to account for the reduced surface tension effects.
Scratch Tip Radius Effects on K-Controlled Slow Crack Growth in PE Pipe
Authors: Corleto Carlos R. , Cole Brian B.
Published: July 2009

A study to evaluate the effect of scratch tip radius on stress intensity factor (K) controlled slow crack growth (SCG) was performed to establish whether a plastics pipe industry practice to allow scratches 10% the thickness of the pipe, could still be allowed on large diameter pipes with blunt scratches. A series of finite element analyses were done using a 1/4 two-dimensional (2-D) notched pipe model assuming a 12-in diameter pipe with a standard dimentional ratio (SDR) of 11, a notch ten times smaller than its thickness, notch tip ratios ranging from 16 to 0.0459, and linear elastic material behavior. Results indicate K-controlled SCG would occur if the ratio of notch tip radius to notch depth is less than 0.1667, although this ratio is probably very conservative due to scratch tip blunting from the formation of a craze zone ahead of crack tips in polyethylene (PE) pipes. However, for ratios greater than 0.5, ductile failures could be induced for internal pressures yielding high hoop stresses and at high temperatures. This is due to the fact that stress concentration factors for relatively blunt notches can still induce maximum scratch tip stresses several times higher than the hoop stress of an unscratched pipe. The results of this finite element analysis could be validated experimentally using ASTM D2837-01 following notching procedures given in ISO 13479 with modified cutters to obtain several notch tip radii.
Mitigation of Stress Corrosion Cracking in Nuclear Weldments Using Low Plasticity Burnishing
Authors: Scheel Jeremy E. , Hornbach Douglas J. , Prevey Paul S.
Published: July 2009

Stress corrosion cracking (SCC) has been observed for decades in austenitic alloy weldments such as type 304 stainless steel as well as in Ni based alloy weldments including Alloy 600 and 690. SCC continues to be a primary maintenance concern for many components in both pressurized water reactors (PWR) and boiling water reactors (BWR). SCC is understood to be the result of a combination of susceptible material, exposure to a corrosive environment, and tensile stress above a threshold. Tensile residual stresses developed by prior machining and welding can accelerate SCC. A surface treatment is needed that can reliably produce deep compressive residual stresses in austenitic and Ni based alloy weldments in order to prevent SCC. Post-weld surface enhancement processing via low plasticity burnishing (LPB) can be used to introduce deep compression into tensile fusion welds thereby mitigating SCC. LPB has been developed as a rapid and inexpensive surface enhancement method adaptable to existing CNC machine tools or robots. Deep compressive residual stresses produced by LPB are designed to reduce the surface, and near surface stress state to well below the SCC threshold. Residual stress results are shown for 304 stainless steel, Alloy 22 and Alloy 718. SCC test results comparing LPB treated and un-treated 304 stainless steel weldments are presented. Results show that the deep compression produced by LPB eliminates SCC in austenitic weldments.
The Development of the Evolutionary BWR (AB1600)
Authors: Murase Akira , Nakamaru Mikihide , Hamazaki Ryoichi , Kuroki Masahiko , Takahashi Munetaka
Published: July 2009

Considering the delay of the first breeding reactor (FBR), it is expected that the light water reactor will still play the main role of the electric power generation in the 2030’s. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to have independent countermeasure for severe accidents and better probability of core damage frequency (CDF) considered external events such as earthquake. On the other hand, we have another goal of the AB1600, which is to retain the safety performance superior or equivalent to the current ABWR without deterioration of economy. In order to achieve both economy and safety performance, we have optimized the safety system configuration of the AB1600 by partly introducing passive safety system to design basis event (DBEs). At the same time, we have adopted the simplification of the overall plant systems in order to improve economy. In order to reduce capital cost, to shorten refueling period and to reduce maintenance effort, the AB1600 introduces the large fuel bundle size. The bundle size is 1.2 times as large as that of the ABWR and the fuel rod array is 12 by 12. And then by progressing the core design, we can reduce the number of reactor internal pumps (RIPs) to eight from the current ABWR of ten. The core power density, the number of fuel bundles, and the core diameter of AB1600 are decided in order to achieve 24 months fuel cycle length on the condition with below 5wt% enrichment of fuel and with eight RIPs.
Development of Decontamination Method for the Concrete Contaminated With C-14 Nuclide in Fugen
Authors: Kitamura Koichi , Hayashi Hirokazu , Shimizu Mitsumori , Matsuo Kiyoshi , Namba Haruyuki , Murakami Kazuo
Published: July 2009

Planning and preparation on decommissioning of the Fugen Nuclear Power Plant have been carried out, and it is one of the most important issues to reduce amount of the radioactive waste for decreasing the disposal cost of them in the dismantling process. In Fugen, the concrete of the reactor building is contaminated with C-14 that was generated due to the activation of CO2 gas during the plant operation. In this study, concrete samples were taken from the reactor building to survey how C-14 is penetrated into the concrete. Concentration of C-14 and C-14 existing rate in each element of the sample specimens were measured in order to investigate C-14 existing feature in the concrete. Besides, depth of the neutralization was measured in order to study on potential relationship with penetration of C-14. Then the following results were obtained: (1) C-14 is penetrated deep beneath of the surface, the depth is approximately as deep as 250mm. (2) C-14 exists only in cement portion of the concrete. It does not exist in aggregates. (3) Analytical study showed that CO2 gas containing C-14 could be penetrated into deeper position in hot and dried air condition during reactor operation, which is coming from that CO2 gas penetrating into concrete tend to move faster in void portion of the porous concrete than in liquid portion of that. Possible methods for removal of C-14 from the concrete are studied based on the above findings. This study showed that C-14 is localized in confined cement portion in 250mm depth of concrete. It is also cleared that CO2 gas containing C-14 in concrete could be removed by heating at a particular temperature. Therefore, it is expected that almost C-14 could be removed from concrete using the combination of both separating of cement portion and heating of it. In the next step of this study, planning the practical procedure for dismantling the reactor building and the removal and disposal method of C-14 will be carried out on the basis of the results of this study. The widely measurement with sampling more specimens is also planned in order to investigate C-14 contaminated distribution of the whole reactor building.
ETDR Project: Investigation of System Effects With the CATHARE Code
Authors: Dor Isabelle , Pignatel Jean-Franc¸ois , Poette Christian , Morin Franck , Bertrand Fre´de´ric
Published: July 2009

The Gas Fast Reactor (GFR) is one of the six concepts selected for further studies in the frame of the GEN IV forum and it is of high priority in the French Atomic Commission (CEA) R&D program on the future Nuclear Energy Systems. The conceptual design is shared between European partners through the GCFR 6th Framework Program. The ETDR (Experimental Technology Demonstrator Reactor) will be the first GFR ever built. The main characteristics of the ETDR ([1]) are a 50 MW concept with a starting core design with existing or close-to-existing technology (MOX pin S/A with metallic cladding) to qualify the whole sub-assembly concept, with an inlet temperature of 260°C and an outlet temperature of 560°C. Under accidental design operating conditions the maximum clad temperature should not exceed 850°C. The system design is largely influenced by the Decay Heat Removal strategy (DHR). In case of depressurised transients, DHR helium flow always relies on forced convection (the back up pressure is assumed equal to 3 bar) whereas in case of pressurised transients, DHR helium flow relies on forced convection but in case of failure of the blower, natural convection should be efficient enough to ensure the core cooling. Therefore a specific system has been designed. It consists of three loops in extension to the vessel including a heat-exchanger, located at a certain elevation in order to ensure natural convection for pressured situation, and a blower in order to start and/or maintain DHR flow especially when low pressures are reached. A CATHARE modelling based on the actual design has been developed and is actually used to carry on first accidental transient analysis on the starting core. The paper first describes the CATHARE modelling. Then the paper analysed two different families of protected transients (pressurised and depressurised transients) and some aggravating events are considered. Finally perspectives and future work are discussed.
Core-Catcher Coolability Analysis With Coupled THEMA and RELAP Codes
Authors: Chen Bin
Published: July 2009

Core-catcher concept is widely used as a mitigation measure for containment protection under severe accident condition. Preliminary design of core-catcher is combined with ex-vessel corium behavior and corium cooling. The loose coupling of THEMA and RELAP codes provides a convenient method to evaluation the coolability of core catcher with a realistic ex-vessel corium characteristic by couple the corium and concrete basemat parameters like spreading area heat flux and etc. calculated by THEMA code and thermal-hydraulic transient behavior of coolant which flows through the basemat and is calculated by RELAP. In this paper the calculation results of coupling methodology is compared with experiment result and simulation of finite element model. The calculation results are consistent to the BENSON test rig, which demonstrates the heat removing capability of EPR core-catcher. However, it also reveals the flow instability in coolant channel, which penalizes the core-catcher coolability. But with an improved design of core-catcher with pre-filling water and core-catcher seal, which we presented in this paper, effect of flow instability can be mitigated, especially for the inlet and outlet of coolant channel.
Seismic Response Analysis of Reactor Coolant Pump in Nuclear Power Plant
Authors: Ren Zhaohui , Ma Hui , Li He , Song Guiqiu , Zhou Wenjian
Published: July 2009

The reactor coolant pump in nuclear power plant is the only revolving equipment in the nuclear power plant. Its functional stability will directly affect the security of nuclear power plant. The coolant pump of a very nuclear plant is examined by using response spectrum analysis to analysis dynamic characteristics and responses aiming at finding the natural frequencies of vibration, modes of vibration and seismic responses, and any possible step which may cause damage of the whole system. The favorable spectrum and unfavorable one are investigated as well. The paper focuses on avoiding the detrimental effects caused by earthquakes, therefore may lay down a theoretical foundation for structural design and installation.
Development of Significance Determination Process in Japan
Authors: Kasagawa Yusuke , Sugawara Masajiro , Uchida Tsuyoshi , Ogura Katsunori , Yamashita Masahiro
Published: July 2009

In September 2006, the regulatory body of Japan, the Nuclear Industry and Safety Agency (NISA), issued an interim report entitled “The improvement of the inspection system for nuclear power plants” which had been reviewed by the Subcommittee of the Advisory Committee on Nuclear and Industrial Safety. The report addresses the potential use of risk information in order to identify the safety significant inspection scope, to select and evaluate performance indicators, to evaluate the safety significance of inspection findings, and to enhance the maintenance program. NISA has been preparing for the new inspection system in Fiscal Year 2008. Before the implementation, technical bases such as the detailed design of the new inspection system and the pilot application of major issues need to be developed. The technical support of this new inspection program is now in progress by the Japan Nuclear Energy Safety Organization (JNES) to develop methodology and technical bases for improvement of efficiency and transparency of regulatory inspection, application of risk information to develop maintenance program guidelines, selection of performance indicators, identification of the safety significance of inspection findings and comprehensive evaluation of individual plants. This paper shows the development and sample calculation of significance determination process (SDP) which is one part of the new inspection program. The SDP is applied to evaluate the significance of inspection findings. The inspection findings are categorized into four groups such as the safety function facet, the risk facet, the public and occupational radiation exposure facet and the safety importance (SI) of the inspection findings are evaluated with risk information. The sample calculation with this SDP indicated that the level of SI is the same level by the current deterministic evaluation process. At present, the SDP models have been developed into the eight types of typical Japanese nuclear power plants for Boiling Water Reactor (BWR), BWR-3, BWR-4, BWR-5, Advanced BWR, 2-Loop, 3-Loop, 4-Loop Dry Containment Pressurized Water Reactor (PWR) and 4-Loop Ice-Condenser PWR.
Numerical Analysis of Heat Transfer Test of Supercritical Water in a Tube Using the Three-Dimensional Two-Fluid Model Code
Authors: Misawa Takeharu , Yoshida Hiroyuki , Tamai Hidesada , Takase Kazuyuki
Published: July 2009

The three-dimensional two-fluid model analysis code ACE-3D is developed in Japan Atomic Energy Agency for the thermal design procedure on two-phase flow thermal-hydraulics of light water-cooled reactors. In order to perform thermal hydraulic analysis of SCWR, ACE-3D is enhanced to supercritical pressure region. As a result, it is confirmed that transient change in subcritical and supercritical pressure region can be simulated smoothly using ACE-3D, that ACE-3D can predict the results of the past heat transfer experiment in the supercritical pressure condition, and that introduction of thermal conductivity effect of the wall restrains fluctuation of wall.
Study on Subcooled Flow Boiling Heat Transfer Under Oscillatory Flow and Vibration Conditions
Authors: Ohtake Hiroyasu , Koizumi Yasuo , Higono Norihumi
Published: July 2009

Onset of nucleate boiling — ONB — and critical heat flux — CHF — on subcooled flow boiling under oscillatory flow and vibration conditions, focusing on liquid velocity, amplitude and frequency of oscillation and vibration were investigated experimentally. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. Frequency of oscillatory flow was 2, 4 and 6 Hz, respectively; amplitude of oscillatory flow was 25 and 50% in a ratio of main flow rate, respectively. For the vibration condition, the test section was set on a vibration table. The acceleration was 1.3 and 4.2 m/s2, respectively; the frequency was 2, 4 and 20 Hz, respectively. The present experimental results showed that temperature at ONB and critical heat flux for oscillatory flow were lower than those for steady flow. The decreasing of liquid velocity by oscillatory caused the ONB and the CHF to decrease. Critical heat fluxes under the vibration conditions were higher than those for steady flow. The CHF under the vibration condition was increased with an increasing of acceleration of vibration. According to present observations, coalesced bubble on the heater was frequently released by vibration of the test heater. This behavior causes the CHF to become higher under the vibration condition. Furthermore, the effects of oscillation and vibration on velocity profile in a channel were estimated to discuss behaviors of a coalesced bubble on a heater.
Development of an ASME Code for Graphite Core Components
Authors: Sanna Christian A.
Published: July 2009

In October 2003, a meeting was held at Oak Ridge National Laboratory to discuss a proposal to restart the development of the draft ASME BPV Section III, Division 2, Subsection CE code. The intent was to update, expand and attain approval of a graphite core components code to support Generation IV nuclear energy projects.
Anticipated Transient Without Scram Analysis for an ABWR Using RETRAN
Authors: Tsai Chiung Wen , Yang Shu Ming , Shih Chunkuan , Wang Jong-Rong , Ma Shao Shih , Cheng Su Chin
Published: July 2009

A RETRAN02/MOD5 model was developed for Lungmen ABWR and applied for ATWS transient analysis. Three ATWS events including Main Steam Isolation Valve Closure (MSIVC), Loss of Offsite Power (LOOP), and Inadvertent Opening of all Turbine Bypass Valves (IOTBV) are analyzed in this study. During the first two transients, the vessel pressure is increased as a result of steam flow reduction due to the closure of main steam isolation valves (MSIVs) and Turbine Control Valves (TCVs) respectively. In the third transient, the vessel pressure is reduced because of the open of Turbine Bypass Valves (TBVs) and turns to be increased because of the closure of MSIVs. All of the above transients suffer high neutron flux as a result of void reduction. There are several equipments and procedures to mitigate ATWS transient such as feedwater trip, Reactor Internal Pumps (RIPs) runback and trip, and the depressurization of relief valves. After the ATWS high pressure signal is initiated and permissive for 180 seconds, Standby Liquid Control system is initiated to inject boron liquid into upper plenum to shutdown the reactor. The results conclude that equipments and procedures mitigate the event effectively and the core is brought to shutdown state.
Measurement of Flow Phenomena in a Lower Plenum Model of a Prismatic Gas-Cooled Reactor
Authors: McIlroy Hugh M. , McEligot Donald M. , Pink Robert J.
Published: July 2009

Mean-velocity-field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined cross flow — with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate geometry scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages in and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3-D) Particle Image Velocimetry (PIV) system was used to collect the data. Inlet jet Reynolds numbers (based on the jet diameter and the time-mean bulk velocity) are approximately 4,300 and 12,400. Uncertainty analyses and a discussion of the standard problem are included. The measurements reveal developing, non-uniform, turbulent flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and presentations that describe the component flows at specific regions in the model. Information on inlet conditions is also presented.
The Error Bounds for Three-Dimensional Nodal LTSN Method
Authors: Hauser Eliete Biasotto , Pazos Ruben Panta , Vilhena Marco T. M. B. , Barros Ricardo C.
Published: July 2009

In this paper we present a proof about the convergence of the 3D Nodal-LTSN Method in order to solve the transport problem in a parallelepiped domain. For that, we define functions associated to the errors, one in the approximated flux, another in the quadrature formula and establish a relation between them. We present a Nodal-LTSN method to generate an analytical solution for discrete ordinates problems in three-dimensional cartesian geometry. We first transverse integrate the SN equations and then we apply the Laplace transform. The essence of this method is the diagonalization of the LTSN transport matrices and the spectral analysis garantees this. The transverse leakage terms that appear in the transverse integrated SN equations are represented by exponential functions with decay constants that depend on the characteristics of the material of the medium the particles leave behind. We present numerical results generated by the offered method applied to typical shielding model problems.
Hardware-in-the-Loop Simulation Using Tricon v9 Safety PLC
Authors: Rankin Drew J. , Jiang Jin
Published: July 2009

This paper presents the performance of shutdown system one (SDS1) implemented on a programmable logic controller (PLC) within real-time hardware-in-the-loop (HIL) simulation. SDS1 evaluation is focused on steam generator (SG) level low trip scenarios. A comparison of the findings with simulated expected plant operation is performed. An Invensys Triconex Tricon v9 safety PLC is interfaced to a real-time nuclear power plant (NPP) simulation suite (DarlSIM), replicating the operation of the Darlington NPP SDS1. Design basis accidents (DBA) associated with SDS1 regulatory standards are developed and applied to the two simulation environments. HIL simulation is a preferred method for testing systems prior to installation and is necessary to ensure proper SDS verification and validation. The performance of the Tricon v9 PLC, the HIL simulation platform and the two simulation environments are evaluated.
Susceptibility of Intergranular Corrosion for Extra High Purity Austenitic Stainless Steel in Nitric Acid
Authors: Ioka Ikuo , Kato Chiaki , Kiuchi Kiyoshi , Nakayama Junpei
Published: July 2009

Austenitic stainless steels suffer intergranular attack in boiling nitric acid with oxidants. The intergranular corrosion is mainly caused by the segregation of impurities to grain. An extra high purity austenitic stainless steel (EHP alloys) was developed with conducting the new multiple refined melting technique in order to suppress the total harmful impurities less than 100ppm. The basically corrosion behavior of type 310 EHP alloy with respect to nitric acid solution with highly oxidizing ions was investigated. The straining, aging and recrystallizing (SAR) treated type 310 EHP alloy showed superior corrosion resistance for intergranular attack. The segregated boron along the grain boundaries was one of main factor of intergranular corrosion from fission track etching results. The SAR treatment was effective to restrain the intergranular attack for type 310 EHP alloy with B less than 7ppm.