bluuebunny/crossref_metadata_2025 huggingface dataset:
On the Use of the Subharmonic Resonance as a Method for Filtration
Authors: Hammad Bashar K. , Nayfeh Ali H. , Abdel-Rahman Eihab M.
Published: April 2011

We study the feasibility of employing subharmonic resonance of order one-half to create a bandpass filter. A filter made up of two clamped-clamped microbeam resonators coupled by a weak beam is employed as a test design. We discretize the distributed-parameter system using the Galerkin procedure to obtain a reduced-order model composed of two nonlinear coupled Ordinary Differentiation Equations (ODEs). It accounts for geometric and electric nonlinearities as well as the coupling between these two fields. Using the method of multiple scales, we determine four first-order nonlinear ODEs describing the amplitudes and phases of the modes. We use these equations to determine closed-form expressions for the static and dynamic deflections of the structure. The basis functions in the discretization are the linear undamped global mode shapes of the unactuated structure. We found that it is impractical to use the proposed filter structure for subharmonic resonance-based filtering since it cannot produce a single-valued response for small excitation amplitudes. On the other hand, it is feasible to use cascaded uncoupled resonators to build a bandpass filter by operating one in the softening domain and the other in the hardening domain.
A Unified Approach for Flow Analysis of Magnetorheological Fluids
Authors: Kavlicoglu Barkan , Gordaninejad Faramarz , Wang Xiaojie
Published: April 2011

This study presents a new approach for flow analysis of magnetorheological (MR) fluids through channels with various surface topologies. Based on an experimental study an analytical method is developed to predict the pressure loss of a MR fluid as a function of the applied magnetic field strength, volumetric flow rate, and surface topology, without utilizing the concept of shear yield stress. A channel flow rheometer with interchangeable channel walls is built to demonstrate that the pressure loss across the MR fluid flow channel is significantly affected by the channel surface properties. Based on the experimental study it is concluded that a unique shear yield stress cannot be defined for a given MR fluid, since its pressure drop depends on the surface topology of the device. Therefore, a relation for nondimensional friction factor associated with MR fluid channel flow is developed in terms of a modified Mason number and dimensionless surface topology parameters. Using the nondimensional model, the pressure loss for various magnetic fields and volumetric flow rates can be represented by a single master curve for a given channel surface topology without the assumption of a constitutive model for MR fluids.
Modal Analysis of the Ice-Structure Interaction Problem
Authors: Venturella Michael A. , Patil Mayuresh J. , McCue Leigh S.
Published: April 2011

In this paper the authors present a multimode ice-structure interaction model based on the single degree of freedom ice-structure interaction model initially proposed by Matlock et al. (1969, “A Model for the Prediction of Ice-Structure Interaction,” Proceedings of the First Offshore Technology Conference, Houston, TX, Vol. 1, pp. 687–694, Paper No. OTC 1066; 1971, “Analytical Model for Ice Structure Interaction,” ASCE Journal of the Engineering Mechanics Division, EM4, pp. 1083–1092). The model created by Matlock et al. assumed that the primary response of the structure would be in its fundamental mode of vibration. In order to glean a greater physical understanding of the ice-structure interaction phenomena, it was critical that this study set out to develop a multimode forced response for the pier when a moving ice floe makes contact at a specific vertical pier location. Modal analysis is used in this study, in which the response of each mode is superposed to find the complete modal response of the entire length of a pier subject to incremental ice loading. This incremental ice loading includes ice fracture points as well as loss of contact between ice and structure. In the work of Matlock et al., the physical system is a bottom supported pier modeled as a cantilever beam. Realistic conditions such as ice accumulation on the pier modeled as a point mass and uncertainties in the ice characteristics are introduced in order to provide a stochastic response. The impact of number of modes in modeling is studied as well as dynamics due to fluctuations of ice impact height as a result of typical tidal fluctuations. A Poincaré based analysis following on the research of Karr et al. (1992, “Nonlinear Dynamic Response of a Simple Ice-Structure Interaction Model,” Proceedings of the 11th International Conference of Offshore Mechanics and Arctic Engineering, Vol. 4, pp. 231–237) is employed to identify any periodic behavior of the low and high velocity ice system responses. Recurrence plotting is also utilized to further define any existing structure of the ice-structure interaction time series for low and high speed ice floes. While the Matlock model on which this research is based is admittedly simplistic, the intention of this work is to provide a foundation for future work using time series analysis and modal analysis on more sophisticated models coupling multiple piers and connecting structure for a comprehensive ice-wind-structural dynamics model.
Results of Field Monitoring on Ice Actions on Conical Structures
Authors: Xu Ning , Yue Qianjin , Qu Yan , Bi Xiangjun , Palmer Andrew
Published: April 2011

Ice-structure interaction plays a central part in determining ice loads and ice-induced vibrations. This is a controversial research issue, and many factors make the problem more complicated. The authors have been monitoring several ice resistant structures in the Bohai Sea for 20 years and have measured ice forces and simultaneously observed ice-structure interaction processes. This paper describes typical physical ice sheet–conical structure interaction processes, field data, and theoretical explanations for different ice conditions and structure dimensions. The conclusions are more widely applicable, and we relate them to field work on ice resistant conical structures in other ice-covered regions. Further work will quantify ice loads on conical structures once the interaction process is understood.
Flow and Pressure Drop Calculation Using Two-Ports
Authors: Elnady T. , Elsaadany S. , Åbom M.
Published: April 2011

Exhaust systems should be carefully designed for different applications. The main objective of an exhaust system is to reduce the engine noise. Maximum noise reduction is usually desired to the limit of a certain backpressure, which is set by the engine manufacturer in order not to deteriorate the engine efficiency. Therefore, a parallel calculation of the flow and pressure drop must be performed. The amount of flow flowing through each element will also affect its acoustic properties. Usually, acoustic and flow calculations are done separately on two different software. This paper describes a new technique that enables both calculations to be done using the same input data on the same platform. Acoustic calculations are usually performed in the frequency domain in the plane wave region using the two-port theory and then the acoustic pressure in the system is solved for using well-known algorithms to handle arbitrary connected two-ports. The stagnation pressure and volume flow can also be calculated using the same algorithm by deriving a flow two-port for each element using the stagnation pressure and the volume flow velocity as the state variables. The proposed theory is first discussed listing the flow matrices for common elements in exhaust elements, and then different systems are analyzed and compared with the measurements.
Optimization Based Identification of the Dynamic Properties of Linearly Viscoelastic Materials Using Vibrating Beam Technique
Authors: Ren Zhiyong , Atalla Noureddine , Ghinet Sebastian
Published: April 2011

Sandwich structures with viscoelastic core and metal face sheets are increasingly used in automotive industry to significantly reduce the amplitude of vibration and noise radiation. Several experimental methods such as dynamic mechanical analysis (DMA) and vibrating beam technique (VBT) are used to characterize the dynamic properties of viscoelastic materials as a function of frequency and temperature. This paper investigates the use of a free-free beam setup, as an alternative solution to the classical clamped-free VBT, for a better control of the effect of boundary conditions on the laminated steel specimen. The new setup is developed in combination with a frequency response function based optimization method, to automatically derive the dynamic properties of viscoelastic core materials and generate their master curves. A solver based on the normal mode superposition method, considering the added mass effect of the impedance head, is used in the cost function of the optimization approach. The sandwich model is based on the Ross–Kerwin–Ungar equation, and the four-parameter fractional derivative model is used in conjunction with the Williams–Landel–Ferry equation to describe the frequency and temperature dependent behavior of the viscoelastic material. The master curves are a direct result of the optimization process. Several applications are described to assess the performance of the present method. In particular, a systematic comparison with both the classical VBT and DMA (when available) is presented.
Criterion for Preventing Self-Loosening of Preloaded Cap Screws Under Transverse Cyclic Excitation
Authors: Yang Xianjie , Nassar Sayed A. , Wu Zhijun
Published: April 2011

In this paper, a novel criterion is developed for preventing the self-loosening of preloaded threaded cap screws under cyclic transverse loading. For a known cyclic excitation, the system parameters are used in the formulation of a closed form solution for the minimum fastener preload required for preventing self-loosening. The effect of several key variables is investigated; this includes bearing and thread friction coefficients, cap screw grip length, thread pitch, material, and cyclic amplitude of the transverse excitation. An experimental setup and test procedure are established. Comparison between the experimental and analytical clamp load variation results shows that the proposed criterion can accurately predict the requirements for preventing self-loosening.
An Analysis of the Cooperative Mechano-Sensitive Feedback Between Intracellular Signaling, Focal Adhesion Development, and Stress Fiber Contractility
Authors: Pathak Amit , McMeeking Robert M. , Evans Anthony G. , Deshpande Vikram S.
Published: April 2011

Cells communicate with their external environment via focal adhesions and generate activation signals that in turn trigger the activity of the intracellular contractile machinery. These signals can be triggered by mechanical loading that gives rise to a cooperative feedback loop among signaling, focal adhesion formation, and cytoskeletal contractility, which in turn equilibrates with the applied mechanical loads. We devise a signaling model that couples stress fiber contractility and mechano-sensitive focal adhesion models to complete this above mentioned feedback loop. The signaling model is based on a biochemical pathway where IP3 molecules are generated when focal adhesions grow. These IP3 molecules diffuse through the cytosol leading to the opening of ion channels that disgorge Ca2+ from the endoplasmic reticulum leading to the activation of the actin/myosin contractile machinery. A simple numerical example is presented where a one-dimensional cell adhered to a rigid substrate is pulled at one end, and the evolution of the stress fiber activation signal, stress fiber concentrations, and focal adhesion distributions are investigated. We demonstrate that while it is sufficient to approximate the activation signal as spatially uniform due to the rapid diffusion of the IP3 through the cytosol, the level of the activation signal is sensitive to the rate of application of the mechanical loads. This suggests that ad hoc signaling models may not be able to capture the mechanical response of cells to a wide range of mechanical loading events.
Probabilistic Safety Assessment to the 1000 MWe NPP During Shutdown Conditions
Authors: Li Lin , Zhao Bo
Published: April 2011

For nuclear power plant, probabilistic safety assessment (PSA) is an effective tool for risk evaluating, risk recognizing and risk managing. This paper initiate a study focused on shutdown safety for NPP as well as the related PSA approaches. Especially, a PSA preliminary analysis and estimation on shutdown operation to the 1000 MWe NPP are performed. The first part of this thesis presents the methods used in the classification of plant operation state (POS), the identification of initiating event and the quantification of initiating event frequency. The second part centers on configuration of shutdown model used in PSA. Based on the design and operation features during shutdown conditions of NPPs, event trees for the shutdown model are constructed and Human Reliability Analysis (HRA) is performed. The last part is the quantification of the integrated shutdown model by means of Risk Spectrum software. System analysis is based on FuQing nuclear power plant PSA report. Component failure data and initiating event frequencies are largely based on EPS900 or NUREG./CR-6144. Overall point estimation of core damage frequency (CDF) during shutdown is assessed; the most important risk contributors of plant operational states, the dominant initiating events and their contribution to CDF are described respectively.
Study on Critical Heat Flux in Narrow Rectangular Channel With Longitudinal Vortex
Authors: Huang Yan-Ping , Huang Jun , Ma Jian , Wang Qiu-Wang
Published: April 2011

Longitudinal Vortex (LV) is produced by Longitudinal Vortex Generators (LVGs) with high heat transfer efficiency and acceptable pressure loss. Due to the relative long influence distance and simple structure, LVGs may be used in narrow channels with flat surface under high temperature and high pressure water medium, in this paper, the critical heat flux (CHF) is one of most important focus. The test channel has the size of 600 mm (length) × 40 mm (width) × 3 mm (height), was used to research the CHF characteristic of CHF affected by LVGs. The test channel is visual in three sides and remains one side for power supply. The LVGs used in the experiments are 14 mm (length) × 2.2 mm (width) × 1.8 mm (height) in dimensions, and periodically mounted on the inner wall of the steel plate. The parameters that are varied during the experiments as follows, system pressure from 0.43 to 0.85 MPa, inlet mass flow flux from 40.2 to 745.7 kg·m−2·s−1, inlet subcooling from 46.8 to 104.2 °C, exit quality from 0.183 to 0.997, surface heat flux from 0.294 to 2.316 MW·m−2. The experiments show that the CHF is improved by 24.3% while the total pressure drop through the test section is improved by 62.9%. The bubble growth and its evolutionary process in narrow rectangular channel with LVGs have been obtained during a short term when the CHF occurs, and it is found that the bubbles have been affected intensely by LV. Based on these experiment data, the growth and aggregation of bubbles have been depressed by LV, the mass, momentum and energy exchange between cold and hot areas in the test section have been strengthened. As a result, the heat transfer enhancement by LV can be explained by the destruction of thermal boundary layer.
Comparison of Severe Accident Results by Using MAAP5 and MAAP4 Codes
Authors: Wang Te-Chuan
Published: April 2011

MAAP5 (Modular Accident Analysis Program Rev. 5.0.0), developed by Fauske & Associates, Inc.’s (FAI) based on the MAAP4 code, is a severe accident analysis code. It is a computer program capable of simulating the response and mitigation actions of light water reactor nuclear power plants (NPPs), including advanced boiling water reactor (ABWR) during severe accident. A specific loss of all core cooling accident sequence, LCLP-PF-R-N, based on Final Safety Analysis Report (FSAR) of Lungmen (ABWR) NPP, was selected as a based case and simulated by the MAAP5 and MAAP4 codes. The MAAP5 and MAAP4 parameter files for Lungmen NPP were established based on Lungmen NPP design data and the MAAP5 and MAAP4 users’ guides. The main severe accident phenomena and the fission product release fractions associated with the LCLP-PF-R-N sequence were simulated. The purpose of this paper is to compare the analysis results of LCLP-PF-R-N sequence calculated by MAAP5 and MAAP4 codes. The two codes give similar results for important phenomena during the accidents, including core uncovery, core support plate failure, debris relocation to the lower plenum, vessel failure, passive flooder opens, containment overpressure protection system (COPS) activation, noble gases and volatile species (like CsI) release to environment, except for the amount of hydrogen production in core. MAAP5 predicts a greater amount of hydrogen production in core than that of MAAP4. This is because MAAP4 predicts earlier reactor pressure vessel (RPV) depressurization than that of MAAP5. That results in earlier steam exhaustion and oxidation reaction termination in core than those of MAAP5. This paper successfully demonstrates the severe accident of Lungmen NPP, and analysis results can provide useful information for the MAAP5 and MAAP4 users.
Integrated Power Recovery Using Markov Modeling
Authors: Rodgers Shawn , Nelson Paul , Betancourt Coral , Kee Ernie , Yilmaz Fatma
Published: April 2011

The solution to a Markov chain modeling electric power supply to critical equipment in a typical 4-loop pressurized water reactor following a Loss of offsite power event is compared with a convolution method. The standard “convolution integral” approach is described, and an alternative methodology based on a Markov model is illustrated.
Codes Comparison and Analysis of Dissimilar Metal Welds: Ultrasonic Testing in Nuclear Power Plants
Authors: Zhong Zhimin , Zhang Goufeng , Yuan Guanghua
Published: April 2011

Dissimilar Metal Welds have been widely used at safe-related pressure vessels and piping in Nuclear Power Plants. Some industry codes have been developed for nuclear power plants, such as ASME BPVC volume III & XI, Germany KTA 3201.3 and 3201.4 code and Russia code of light water nuclear power plants. The difference of those codes and some industry feed backs and some experiment results have been briefly introduced and discussed. Furthermore, the inspection qualification or performance demonstration, one of new requirements of some codes, especially for pre-service and inservice inspection of nuclear power plants, really promote continual improvements of DMWs UT both on detection rate and sizing accuracy. Additional, these works really benefit the revision of EJ/T 1039-1996, Non-destructive testing for mechanical components in nuclear island of nuclear power plants, as china building more and more nuclear power plants.
Developing Software Configuration Management System for Digital Instrumentation and Control (DI&C) System of Nuclear Power Plant
Authors: Chou I-Hsin , Hsiao Hsueh-Wei , Chang Che-An
Published: April 2011

As Nuclear Power Plant (NPP) adopted software-based Digital Instrumentation and Control (DI&C) System, the Software Configuration Management (SCM) is becoming more and more important for NPP. The main reason is the inherent changeability and invisibility which of software often causes unpredictable problems and results are difficult to manage. In addition, the DI&C system has always been constructed by multiple vendors and each vendor of the DI&C has its own development artifacts. Therefore, it is great challenge for NPP staff to maintain the consistency and integrity of software Configuration Items (CI) among multiple vendors. The software CIs include software products delivered to customers and items required to create the software products such as software design document, source code, database, test report, compiler, etc. In general, Software Configuration Management System (SCMS) is usually developed to support SCM activities, such as storing CI, controlling change, and accounting and auditing throughout the entire software lifecycle. However, most existing file-based SCMS typically deal with those artifacts of individual files without providing the more detailed configuration and change information among CIs. Based on the nuclear SCM related regulations, this paper proposes a developing SCMS for the DI&C system of a NPP. Its main goal is to meet the regulatory requirements, and enhance the visibility, tractability and integrate ability to manage the heterogeneous subsystems within the DI&C system. This paper provides the more detailed descriptions about regulation requirements analysis, system design and the development process. Finally, a prototype system is presented.
Flow Instability Study in Rectangular Channel
Authors: Lu Donghua
Published: April 2011

The density wave oscillation (DWO) was investigated with parallel rectangular channel test sections, which have 2mm*25mm flow cross section. Test parameters are pressure 1MPa–10MPa, mass velocity 200kg/m2s–800kg/m2, and inlet subcooling 10°C–50°C. The experimental data show that heat flux rises with high mass velocity, inlet subcooling or system pressure at the stable/unstable boundary. In them, increasing mass velocity can greatly improve stability of this two parallel channel system. Period of the oscillation becomes shorter while mass velocity rises, but when inlet subcooling is increased, it becomes longer. The comparison between data from rectangular channels and round tubes indicates they have the same trend based on the dimensionless phase change number and subcooling number.
Simulation of Thermal-Hydraulic Process of Reactor in HTR-PM
Authors: Zhou Kefeng , Zhou Yangping , Ma Yuanle , Sui Zhe , Li Fu , Dong Yujie
Published: April 2011

This paper provides a discussion of the model development status and verifications for the components used in the Reactor Thermal-Hydraulic model for the Full-Scale Simulator (FSS) of the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). Heat transfer network and thermal-hydraulic network are adopted to simulate the thermal-hydraulic process of reactor of HTR-PM. Due to the First of a Kind Engineering nature and lack of reference plant data, model verification will mainly focus on benchmarking the model configurations against test cases performed by HTR-PM design code THERMIX in future.
Fast Breeder Reactor Core Concept to Improve Nuclear Proliferation Resistance
Authors: Fujimura Koji , Itooka Satoshi , Nitawaki Takeshi
Published: April 2011

A sodium-cooled MOX-fueled FBR core concept to improve nuclear proliferation resistance was proposed. First, we set an index for the nuclear proliferation resistance. In a previous study, reactor-grade Pu was defined such that the Pu-240 isotopic ratio was larger than 18%. Another study defined nuclear proliferation resistance with the Pu-238 isotopic ratio considering its higher spontaneous fission rate and decay heat. We tentatively use the total isotope composition ratio of Pu-238 and Pu-240 as a proliferation resistance index in line with the earlier studies. Next, we designed the sodium-cooled mixed-oxide (MOX)-fueled core concept with the breeding ratio (BR) of over 1.1 without a radial blanket. To attain the index for nuclear proliferation resistance, we added minor actinides (MAs) to the axial blanket fuel (AB). Contents of MAs in the AB to achieve the proliferation resistance index were evaluated. For the case of Np as a representative MA, the minimum content of Np to achieve the index was 3%. And, for the case of loading all MAs, the minimum content of MAs was 10.5%.
Simulation and Realization of Nuclear Instrument System
Authors: Zuo Guangxia , He Bin , Ma Wenyan , Qiu Xiaolin , Hui Weihua
Published: April 2011

For the special nature of the nuclear physics experiment, a NIM-standard nuclear instrumentation system is simulated based on LabVIEW, including plug-ins simulation and integrated devices such as a nuclear random signal generator, a linear amplifier, a single channel analyzer, a scaler and a data-processing module. In order to verify the system, some undergraduate nuclear physics experiments were done. The results show that the simulation signals of the nuclear random signal generator accord with the rule of statistical nature of the nuclear random pulse, the GUI of simulation system is verisimilar, simulation operation is basically the same as real experimental operation. Application of the simulation system has practical significance in mastering conventional nuclear radiation detection methods and avoiding radiation damage of the experiments.
Life Extension of the Fast Breeder Test Reactor
Authors: Anandapadmanaban B. , Babu A. , Babu B. , Dinesh K. , Ramanathan V. , Srinivasan G.
Published: April 2011

The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam, India. The reactor went critical in October, 1985 with a core of 23 unique high Plutonium carbide fuel subassemblies and the reactor power was rated for 10.5MWt with peak linear heat rating of fuel at 320W/cm. The extension of the target burn-up of this fuel based on Post Irradiation Examination at different stages enabled progressive expansion of the core and increase in reactor power. The reactor has been operated upto a power level of 18.6MWt/3MWe with a sodium temperature of 482°C max. The reactor has completed 24 years of operation and is currently under periodic safety review by the Atomic Energy Regulatory Board of India. As a part of the periodic safety review, equipment qualification status and ageing management studies have been presented to the regulators. Equipment qualification refers to the ability of the replaceable equipment to meet the functional requirements on demand, accomplished by periodic surveillance, maintenance and replacement. Ageing management addresses the residual life assessment of components which are passive, non-replaceable / replaceable with difficulty, taking into account their life degrading mechanisms. Over a period of time, based on the operational feedback, maintenance difficulties and obsolescence, several major components have been replaced. These include the Neutronic channels, UPS, computers of the Central Data Processing System, main boiler feed pumps, three control rod drive mechanisms, two control rods, central canal plug, deaerator lift pumps, reheaters of the steam water system, station batteries, DM plant and Nitrogen plant. The starting air system of the emergency diesel generators and isolation dampers of the reactor containment building have also been replaced. Regarding the non-replaceable components, residual life assessment has been carried out based on the operational history vis-a`-vis the design limits for each component. The life limiting mechanism of heat transport systems of FBTR are creep and fatigue. Since the reactor has operated only upto a temperature of 444°C till 2007, the creep effect is insignificant. The total number of thermal cycles seen by the reactor components as of 2007 was 163, as against the design cycle of 2000 for most of the components. Hence all the heat transport system components are as good as fresh ones. However, the major life limiting factor has been found to be the Neutronic fluence on the grid plate which supports the core. The fast flux at the grid plate location was measured using Np foils and the residual life of the reactor has been assessed to be 10.5 effective full power years. This paper details the life extension exercise being carried out for FBTR.
Neutronics Analysis of the Target-Moderator-Reflector (TMR) Configuration for Compact Long-Pulse Neutron Sources
Authors: Zhong B. , Liang T. J. , Zha H. , Yu G. L. , Wang K. , Wei J. , Loong C.-K.
Published: April 2011

The compact pulsed neutron source facility can play an important role in the research, education, user training, and development of the advanced neutron scattering instruments. The materials for the target, moderator, reflector (TMR) and their configurations must be optimized to get the optimal yield of neutrons with energy in the range of 1 meV to eV order which depends on the proton energy and its nuclear reaction. Several kinds of materials of the TMR, their configurations, and their dimensions are investigated by the Monte Carlo simulation and optimized for developing the compact pulsed neutron source. The results would contribute to the construction of the Compact Pulsed Hadron Source (CPHS) of Tsinghua University.
Experimental Study of Temperature Sensitive Dyes for Planar Laser Induced Fluorescence Thermometer
Authors: Tan Sichao , Estrada-Perez Carlos E. , Dominguez-Ontiveros Elvis E. , Hassan Yassin A.
Published: April 2011

As a non-intrusive, whole-field temperature measurement technique, LIF (Laser Induced Fluorescence) has been used successfully to measure temperature fields. The performances of dyes are essential of the technique, especially the temperature sensitivity of the dyes. This work presents an analysis to provide a correct choice of temperature sensitive dyes combination (FL27 and RhB). The influences of temperature, excitation wavelength and pH on emission intensity and temperature sensitivity were analyzed. The results show that the temperature dependent tendency of FL27 changed from negative to positive as the excitation wavelength increased. The temperature sensitivity (4.0% per °C) of combination under 532nm laser is better than that of the wide used combination of RhB and Rh110 (2.0% per °C). The emission intensities of dyes are sensitive to pH value; however, the temperature dependence is unaffected.
A Methodology for Updating Living Probabilistic Safety Assessment
Authors: Zubair Muhammad , Zhang Zhijian , Khan Salah Ud-din , Hua Sheng Den
Published: April 2011

After the construction of nuclear power plants (NPPs) the main aim is to achieve high level of safety. For this purpose different methods and techniques such as defense in depth, safety culture, human reliability analysis (HRA), human factor engineering (HFE), fault tree analysis (FTA), event tree analysis (ETA), deterministic safety analysis (DSA), and probabilistic safety assessment (PSA) etc. have been used for many years and also in present days. Although these methods are suitable for safety of NPPs but with the passage of time, changes occur in components reliability and operating procedures, which continuously modify configuration of NPP. In order to handle these situations, living probabilistic safety assessment (LPSA) and risk monitoring (RM), as an application of PSA, play an important role in updating and maintaining level of safety. The objective of this paper is to summarize the history of LPSA and RM. The study also highlights a newly developed risk monitor called Risk Manager that has been presented in this paper.
Numerical Evaluation of Fluctuating Pressure at Stub Pipes in Actual BWR Main Steam Lines
Authors: Morita Ryo , Takahashi Shiro , Inada Fumio , Ogawa Yukio
Published: April 2011

The power uprating of the nuclear power plant is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. We have started to develop the evaluation methods of pressure fluctuations at the dryer in BWR-5, which are the main type of BWR in Japan. In this study, to investigate the fluctuating pressure at rated power condition, we conducted steam flow simulations in actual main steam piping geometry and conditions. As a result, small amplitude resonance was observed in rated power condition.
Results of Researches in the Frame Work of the Project “OECD PSB-VVER”
Authors: Melikhov Oleg I. , Elkin Ilya V. , Melikhov Vladimir I. , Nikonov Sergey M. , Pylev Sergey S.
Published: April 2011

Under OECD PSB-VVER project, five experiments were performed in the PSB-VVER test facility. They are: 11% upper plenum break, natural circulation test, small break in cold leg, leak from primary circuit into secondary one and large break in cold leg. The experimental program purposes to cover the conditions that are highly relevant to the code validation related to PWR in general and to VVER-1000 safety assessments in particular. Main results of the OECD PSB-VVER project are presented in the paper.
Compact Engineering Simulator for HTR-PM by Embedding THERMIX Code in vPower Simulation Platform
Authors: Gao Qiang , Zhou Yangping , Zhou Zhiwei , Sui Zhe , Ma Yuanle , Li Fu
Published: April 2011

This paper describes the development of a compact engineering simulator of Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) by embedding THERMIX code into the vPower simulation environment. The compact engineering simulator consists of modules for two reactors, two steam generators and entire secondary loop system for power generation with a water-steam Rankin cycle. Two THERMIX modules are employed to simulate the two primary loops corresponding to the two reactors in the HTR-PM respectively. Then, the vPower synchronizes the two THERMIX modules in executing simulation to such two reactor module-structures of HTR-PM. The simulation modules for secondary loop and human machine interface are mainly developed with intrinsic models of vPower simulation platform. Current simulation results are in good agreement with the design values and safety analysis results of HTR-PM. The compact HTR-PM engineering simulator will be improved and validated for safe analysis, procedure development and design change verification.
Design of ITER TF Magnet Cryostat Feeder Through
Authors: Zhou T. , Song Y.
Published: April 2011

International Thermonuclear Experimental Reactor (ITER) TF feeder systems convey the cryogenic supply and electrical power to the TF coils. The Cryostat Feed-through (CFT) includes the straight feeder part from the cryostat wall to the S-Bend Box (SBB). It is the bottleneck of the feeders. The huge Lorentz-force is a challenge for the CFT design. So the reasonable distribution and structural design of the internal and external supports are important. The CFT include the cold (cryogenic) to warm (room temperature) transitions. It is highly integrated with the cryo-pipes, the busbars, the superconductor joints, the thermal radiation shield and the instrumentation pipes and so on. The cryogenic and electrical requirements, the vacuum and mechanical requirements, and so on are considered when the CFT is designed. This paper presents the functional requirements on the TF CFT, gives its structure. The supports are designed and arrayed according to their mechanical or thermal function separately to stand the huge mechanical loads and isolate the conducting heat load from room temperature respectively. The assembly scheme is also described. Mid-joint and cryostat joint are designed to give the facility for the assembly on location. The mechanical analysis result shows the stress in the stainless steel and G10 material both are within the materials stress safety margin. The heat load to the cryogenic pipes and busbars are also less than the requirement 15W. Transient thermal analysis of global feeder model indicates that 32 days are needed for the feeder components to cool down to the required condition.
Decay Heat Benchmark for Uranium-Free Fuels With Minor Actinides
Authors: Rineiski Andrei , Rimpault Ge´rald , Glinatsis Georgios , Messaoudi Nadia , Pelloni Sandro , Schwenk-Ferrero Aleksandra , Vicente Maria Carmen
Published: April 2011

The decay heat (energy due to decay of unstable nuclei) is a small fraction of reactor power at nominal conditions, but after reactor shut-down it is the most important heat source. For taking this source into account in design and safety studies, recommendations are available for fuels of operating reactors, such as UOX and MOX. Fuels for EFIT (European Facility for Industrial Transmutation), unlike UOX and MOX, should contain a significant amount of Minor Actinides (MAs) that would influence decay heat. CEA, CIEMAT, ENEA, FZK (now KIT), PSI and SCK•CEN established a benchmark case and computed decay heat curves for MA-bearing fuels and a MOX-type fuel. The decay heat in the fuels with MAs is appreciably higher than in MOX, except for low burnup cases after short cooling times. This should be taken into account in the design of the decay heat removal system for EFIT. The obtained differences between the decay heat in MA-bearing and MOX fuels are supposed to be representative for the benchmark (or similar) conditions. More effort is needed to evaluate the uncertainties of the computed results.
Qualification of Reactor Physics Toolset for a Thorium-Fuelled CANDU Reactor
Authors: Boczar Peter G. , Hyland Bronwyn , Hagberg Erik
Published: April 2011

AECL has developed enhanced versions of the reactor physics computer codes for analysis of CANDU® reactors and the ACR-1000™. The central codes that comprise the analysis toolset are WIMS-AECL (a lattice code), RFSP (a core code) and MCNP5 (a Monte Carlo code). The toolset, with ENDF/B-VI nuclear data, has been validated for application to the ACR-1000 design. In addition to comparisons of code predictions against relevant experiments conducted in AECL’s ZED-2 critical facility, advanced methods based on cross-section sensitivity/uncertainty (S/U) analysis were used to extend the results of bias and uncertainty in reactivity coefficients, derived from analysis of ZED-2 tests, to the ACR-1000 reactor. The validation of this toolset with ENDF/B-VII nuclear data is proposed for application to analysis of a Thorium-fuelled CANDU Reactor (TCR). The TCR is based on the Enhanced CANDU 6™ (EC6™) reactor [1] and would operate with a fuel design that incorporates both low-enriched uranium (LEU) oxide and thorium oxide fuel elements in the same fuel bundle to achieve enhanced fuel and core performance with thorium fuel. For the initial TCR toolset qualification, important reactor physics phenomena would be validated using several relevant ZED-2 experiments performed in the past. Results from experiments with a variety of oxide fuels are available, including plutonium/thorium (Pu/Th), 233U/Th, 235U/Th, LEU and CANDU-MOX (containing a mixture of plutonium, uranium and dysprosium to simulate the reactor physics affects of fuel burnup). Taken together along with other relevant experimental data, these experiments would be expected to address the important isotopes and many phenomena for the TCR and to enable the validation of the reactor physics toolset for this design. Additional confirmatory experiments would reduce uncertainties. This paper describes the qualification process, including validation, which is proposed to support the use of the reactor physics toolset for the TCR.
Simulation of Heavy Ion Source-Drain Penetration Effect in SOI MOS and Bulk MOS
Authors: Li Yonghong , He Chaohui , Xia Chunmei
Published: April 2011

The Single Event Effects (SEE) of Silicon On Insulator (SOI) and bulk-silicon NMOS are simulated using the SENTAURUS-TCAD device simulator. The Source-Drain Penetration Effect, which is caused by a heavy ion, was shown. It is proved that when the feature size of device become less than a certain scale, both Direct Channel Effect and Indirect Channel Effect occur. By comparing the distributions of equipotential lines in the MOSFETs’ channels of the different feature size devices during the ion strikes indirectly, the Source-Drain-Penetration Effect occurs more evidently when the device feature size is getting smaller.
Effect of Hydrogen Isotope and Concentration on Delayed Hydride Crack Growth Rates in Zr-2.5Nb Pressure Tubes
Authors: Shek Gordon K. , Seahra Harry
Published: April 2011

CANDU Zr-2.5 Nb pressure tubes are susceptible to a cracking mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation and fracture at a crack tip. As defense-in-depth, when DHC is postulated to have initiated from a flaw, it is required to demonstrate that the crack can be detected by the leak monitoring system and the reactor safely shut down before the crack reaches the critical length for pressure tube rupture. DHC growth rates (DHCR) in the axial direction of the tube are required for such leak-before-break assessment. In this test program, the effect of hydrogen isotope and its concentration on DHCR in an unirradiated Zr-2.5 Nb pressure tube is studied. Pressure tube sections were hydrided or deuterided to different concentrations (nominal concentrations of 60, 100 and 190 ppm by weight). For the deuterided tube sections, they contained about 10 ppm of hydrogen from the manufacturing process. The DHC growth rate tests were performed on fatigue pre-cracked curved compact tension specimens, machined from the hydrided or deuterided tube sections, in several stepper-motor controlled load frames with cracking being monitored by direct current potential drop and acoustic emission techniques. DHCR at three test temperatures (270°C, 240°C and 200°C) were obtained from each specimen with the test temperatures approached from a peak temperature of 330°C. Some specimens were tested with a peak temperature of either 370°C or 300°C. The two main conclusions from the study are: (1) DHCR are affected by the hydrogen in solution at the test temperature and not by the amount of bulk hydrides present. The hydrogen in solution at a given test temperature depends on the hydrogen concentration of the specimen, as well as the thermal history (peak temperature in the initial thermal cycle and the test temperature) as a result of the hysteresis of Terminal Solid Solubility between hydride dissolution during heating and precipitation during cooling. (2) The DHC growth rates of the hydrided material are higher than those of the deuterided material because of the higher diffusion rate of hydrogen than deuterium. The isotope effect of hydrogen on DHC growth rates depends on the test temperature, with no apparent effect at 200°C and about 37% difference at 270°C which is slightly below the factor of √2 expected from the mass law of diffusion. The observed temperature dependence could be due to the presence of about 10 ppm hydrogen in the deuterided specimens, which dominates the DHC process at 200°C but insufficient to have a large effect at 270°C. The implication of the observed isotope and concentration effect of hydrogen on DHC growth rates on leak-fore-break assessment of flaws in pressure tubes is discussed.
Improve Safety Lower Cost: Demonstrate Nuclear Design Technology With DOE’s Next Generation Nuclear Plant (NGNP)
Authors: August J. K. , Hunter J. J.
Published: April 2011

Complex nuclear plants have high costs. Documenting and maintaining information contributes a large share of those costs. Improving methods to design, construct and operate nuclear plants lowers costs. Improving design processes could help manage nuclear project costs, birth-to-death. Improving methods that manage design basis information before, during and after construction would not only lower initial nuclear costs, but also provide more complete, accurate information for operations over plant life. Nuclear safety would benefit. The Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) offers a unique opportunity to advance nuclear design methods to improve cost performance. Better design methods could automatically track changes to critical design content reducing laborious effort to update designs, speeding development. They would enhance information sharing. The commercial NGNP project is a significant opportunity to improve commercial nuclear plant safety design. The U.S. DOE should promote innovative nuclear design processes with the NGNP. Providing a single, consistent design framework on a modern data platform would make the NGNP the nuclear industry’s “Go to the Moon” project.
Pre-Test Analysis of Natural Circulation Test of PHENIX End-of-Life With the MARS-LMR Code
Authors: Jeong Hae-Yong , Ha Kwi-Seok , Lee Kwi-Lim , Kwon Young-Min , Chang Won-Pyo , Suk Su-Dong , Kim Yeong-Il
Published: April 2011

PHENIX, a prototype sodium-cooled fast reactor (SFR), has demonstrated a fast breeder reactor technology and also achieved its important role as an irradiation facility for innovative fuels and materials. In 2009 PHENIX reached its final shutdown and the CEA launched a PHENIX end-of-life (EOL) test program, which provided a unique opportunity to validate an SFR system analysis code. The Korea Atomic Energy Research Institute (KAERI) joined this program to evaluate the capability and limitation of the MARS-LMR code, which will be used as a basic tool for the design and analysis of future SFRs in Korea. For this purpose, pre-test analyses of PHENIX EOL natural circulation tests have been performed and one-dimensional thermal-hydraulic behaviors for these tests have been analyzed. The natural circulation test was initiated by the decrease of heat removal through steam generators (SGs). This resulted in the increase of intermediate heat exchanger (IHX) secondary inlet temperature, followed by a manual reactor scram and the decrease of secondary pump speed. After that, the primary flow rate was also controlled by the manual trip of three primary pumps. For the pre-test analysis the Phenix primary system and IHXs were nodalized into several volumes. Total 981 subassemblies in the core were modeled and they were divided into 7 flow channels. The active 4 IHXs were modeled independently to investigate the change of flow into each IHX. The cold pool was modeled by two axial nodes having 5 and 6 sub-volumes, respectively. The reactor vessel cooling system was modeled to match the flow balance in the primary system. The flow path of vessel cooling system was quite complicated. However, it is simplified in the modeling. For a MARS-LMR simulation, the dryout of SGs have been described by the use of the boundary conditions for IHTS as a form of time-to-temperature table. This boundary condition reflects the increase in IHTS temperature by SG dryout during the initial stage of the transient and the increase in heat removal by the opening of the two SG containments at 3 hours after the initiation of the transient. Through the comparison of the pre-analysis results with the prediction by other computer codes, it is found that the MARS-LMR code predicts natural circulation phenomena in a sodium system in a reasonable manner. The final analysis for validation of the code against the test data will be followed with an improved modeling in near future.
Poolside Examination Techniques Applied for Development of an Advanced PWR Fuel, PLUS7™
Authors: Shin Jung-Cheol , Kim Jae-Ik , Kwon Jung-Tack
Published: April 2011

The poolside examination is demanded essentially for the development of an advanced fuel, because it provides data about in-reactor fuel behaviors as well as product performance. Therefore, all the results of examinations are transmitted back to design groups, and evaluated with the design models and codes applied new fuel development. In general, PSE is performed a variety of examinations in assembly and single rod state, such as visual inspection, fuel assembly length growth, bowing and twist, rod-to-rod spacing, spacer grid width, fuel rod diameter, and fuel rod oxide thickness at the end of each cycle of irradiation during plant outage periods. After the completion of fuel life, selected rods are extracted from the fuel assembly for individual fuel rod measurement. In this paper, the techniques of PSE applied an advanced pressurized water reactor fuel, PLUS7™ for Optimized Power Reactors (OPR 1000s) in Korea are described.