bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Schedule Control in AP1000 Nuclear Plant Project
Authors: Fang Xiaopeng
Published: April 2011

The construction of AP1000 Nuclear Power Plant (NPP) has commenced in China. The AP1000 NPP features a passive design concept and modular construction technology. Based on the management of the construction of current AP1000 NNP, this paper describes the effects on Nuclear Island (NI) construction project management resulting from modular construction technology, as well as new construction techniques and methods. This paper puts forward new requirements for construction schedule management of the nuclear island construction at different levels. The AP1000 NI construction logic features the parallel construction of civil and structural erection as the main approach, with the integrated schedule of module fabrication, assembly and installation as support.
Validations of CFD Code for Density-Gradient Driven Air Ingress Stratified Flow
Authors: Oh Chang H. , Kim Eung S.
Published: April 2011

Air ingress into a very high temperature gas-cooled reactor (VHTR) is an important phenomenon to consider because the air oxidizes the reactor core and lower plenum where the graphite structure supports the core region in the gas turbine modular helium reactor (GT-MHR) design, thus jeopardizing the reactor’s safety. Validating the computational fluid dynamics (CFD) code used to analyze the air ingress phenomena is therefore an essential part of the safety analysis and the ultimate computation required for licensing. An experimental data set collected by ETH Zurich on a lock exchange experiment (Grobelbauser et al., Lowe et al. 2002; Lowe et al. 2005; and Shin et al. 2004) was selected for the validation. The experiment was based on a series of lock exchange flows with gases of different density ratios varying from 0.046 to 0.9 in a closed channel of a square cross-section. The focus was on the quantitative measurement of front velocities of the gravity current flows. The experiment results cover the full range of gas intrusions—heavy as well as light—for the gravity current flows in the lock exchange situations. FLUENT CFD code (ANSYS Fluent 2008) was used. The calculated results showed very good agreement with the experimental data. A number of tables and comparison plots are included to summarize the estimated current speeds. The current speed obtained by experimental data was 1.25 m/s and that of the simulation was 1.19 m/s. This result indicates that the deviation of the simulation is only 4.8% that of the experimental data.
Study of Heat Transfer Model of the Spiral-Fin Fuel Element
Authors: Zhang Dan , Liu Changwen , Lu Jianchao
Published: April 2011

Spiral-fin fuel elements belong to the self-spacer fuel elements; they are usually arrayed in hexagonal fuel assembly and widely used in the core of the tight lattice reactor. Closely packed and helical spacer makes the heat transfer course and phenomena in the channel of the spiral-fin fuel bundle very complicated, with little literature being found in this field. The paper preliminarily studied the heat conduction through the finned can together with convection heat transfer between the fuel element and coolant. Modified fin efficiency was introduced to calculate the heat conduction; for the convention heat transfer, an imaginary channel named equivalent annulus was adopted. The geometric and physical influence of the spiral fins to the coolant in the annulus was studied. Comparing with annulus have inner surface being heated, it was found that the heated perimeter and velocity of coolant in the equivalent annulus were changed by the helical spacer, a new model correlation based on the equivalent annulus’ method was obtained by modifying the heat transfer correlation in annulus. Quantitative study was conducted for a given rods bundle, the results shows that the spiral-fins’ effect to the heat transfer is related to the change of the convection coefficient and heat conduction amount caused by the spacer, at the common condition, its impact is little.
The Research on Scale of Cluster Fuel Assembly Spacer Grids
Authors: Chen Jie , Chen Bingde , Zhang Hong
Published: April 2011

The existence of boundary walls will impact flow and heat transfer to some extent. The boundary walls decrease the lateral velocities and change the flow direction to form some eddies, so the flow instability increases. Besides, they stop the mass transfer between the inner areas and outside. For the spacer grids of cluster fuel assembly, in order to confirm the scale of simulation object or experiment noumenon, the first thing is to research the influence depth of boundary effect, and try to eliminate its influence to the concerned object or area. This paper is focused on the spacer grids of 7×7 scale, 5×5 scale and 3×3 scale, respectively. Computed with CFD method, the lateral velocities of corner areas and typical subchannels are compared. And then, the influence depth of boundary effect is confirmed as the distance of a range of fuel rods. Therefore, for the concerned subchannel of CFD simulation, there must be no less than a range of fuel rods in the periphery to mitigate, so the result is not affected by the boundary effect.
A Criterion for the Onset of Air Ingress in VHTR by Air Intrusion
Authors: Liou Jim C. P. , Parks Donald L. , Schultz Richard R.
Published: April 2011

The projected prismatic and pebble-bed preconceptual designs for the Very High Temperature Reactor (VHTR) both have large diameter, horizontally-oriented, exit ducts that lead hot helium from the core exit through the lower plenum to the balance-of-plant. In the event a significant leak is present in the exit duct the system will depressurize and the possibility of air ingress into first the lower plenum and then the core must be considered. The intrusion of air against helium flow is driven by the weight difference between the two fluids. Thus the potential for air intrusion always exists. Whether such a phenomenon will actually occur is not as clear. This paper proposes a simple criterion for the onset of air intrusion. The criterion is expressed in terms of a densimetric Froude number (the inverse of the bulk Richardson number) and a scaled pressure difference between the duct exit and the containment. The pressure difference is rendered dimensionless by dividing it with the product of the duct radius and the difference of specific weight of the two fluids. The dimensionless variables in the criterion enable its validity be verified by experiments using fluids other than helium and air. This paper presents the development of this criterion and experimental validation using water and biodiesel as the testing fluids. The relevance and application of this criterion to VHTR air ingress scenarios are described.
Automating a Nuclear New-Build Field Welding Program to Include Construction Code Rules and Requirements
Authors: Cannell Gary R. , Arms Evan , Chaubey Nikhil
Published: April 2011

Fluor, a large Engineering, Procurement and Construction contractor, recently renewed its ASME certifications for the construction of nuclear power plants. In preparation for a resurgence of commercial nuclear power plant construction, Fluor Nuclear Power (FNP) Construction Welding Engineers have prepared an electronic field welding program, to be used in conjunction with an automated system for the generation, control and documentation of work packages. The prior generation of nuclear power plants constructed in the US utilized a manual process for controlling field welding activities. The “manual” way of doing business required a relatively large, on-site staff (both technical and administrative) to create, issue, track, and document this work. In addition, the manual process was prone to human error. In an effort to improve this key construction activity and reduce construction costs, the FNP Construction Welding group has prepared an electronic welding program that automatically performs the majority of the work package preparation/documentation tasks previously performed manually. The electronic welding system has been designed to access engineering and construction code information related to welding, process the data through a series of logic-based spreadsheets and automatically populate the work package with welding requirements — preheat, post-weld heat treatment (PWHT), Welding Procedure Specification (WPS), etc. The spreadsheets analyze the engineering data (i.e., base material type, thickness, applicable code, joint design, etc.), in conjunction with construction code rules and Fluor welding practices, to determine appropriate welding requirements. System generated requirements are then automatically entered into the work package. This paper describes the design of the electronic welding program, it’s scope, development and qualification. In addition, preparation and qualification of the spreadsheet logic, that effectively translates specified code welding criteria into work package requirements, will be reviewed. It is believed that this type of system will be needed to successfully construct the next generation of US nuclear power plants.
RELAP Simulation on Boiling and Flow Phenomena Under External Reactor Vessel Cooling Condition
Authors: Li Yongchun , Zhou Weihua , Yang Yanhua , Kuang Bo , Cheng Xu
Published: April 2011

External reactor vessel cooling (ERVC) of the In-vessel retention (IVR) system is widely accepted as a feasible way to remove decay heat from the lower head of the reactor pressure vessel (RPV) under severe accident (SA) conditions. However, some issues relating to ERVC still need to be evaluated before its application, such as boiling and flow phenomena and CHF prediction, etc. To study these key issues, an experimental study program named REPEC (Reactor Pressure Vessel External Cooling) is performed at Shanghai Jiao Tong University. Steady state experiments focusing on flow boiling phenomena investigation are carried out with comprehensive measurements, including temperature distribution, pressure drop and mass flow rate. As a part of studies on boiling mechanism and flow phenomena between RPV and the insulation, the experiment is analyzed and simulated with RELAP code. The code simulation covers most of the experimental cases, and a comparison between simulation results and experimental data are presented and discussed.
Thermal-Hydraulic R&Ds for the APR+ Developments in Korea
Authors: Song Chul-Hwa , Kwon Tae-Soon , Yun Byong-Jo , Choi Ki-Yong , Kim Hwan-Yeol , Jun Hyung-Gil , Kim Hang-Gon
Published: April 2011

This paper briefly introduces recent progress in thermal-hydraulic R&Ds, which is mainly being performed at KAERI, for the APR+ (Advanced Power Reactor plus) development. The main R&D items for the APR+ reactor are associated directly with recent efforts to introduce new safety concepts in the APR+ standard design developments, which is currently in progress in the Republic of Korea. The R&D activities reported here mainly cover the thermal-hydraulic and severe accident areas and are being performed in experimental and/or analytical ways. They include: (1) advancement and optimization of safety injection system, (2) incorporation of passive safety features, such as advanced Fluidic Device (FD+) and passive auxiliary feedwater system (PAFS), and (3) incorporation of severe accident mitigation features.
Detailed Analyses of Specific Phenomena in Core Disruptive Accidents of Sodium-Cooled Fast Reactors by the COMPASS Code
Authors: Morita Koji , Zhang Shuai , Arima Tatsumi , Koshizuka Seiichi , Tobita Yoshiharu , Yamano Hidemasa , Ito Takahiro , Shirakawa Noriyuki , Inoue Fusao , Yugo Hiroaki , Naitoh Masanori , Okada Hidetoshi , Yamamoto Yuichi , Himi Masashi , Hirano Etsujo , Shimizu Sensuke , Oue Masaya
Published: April 2011

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include 1) fuel pin failure and disruption, 2) molten pool boiling, 3) melt freezing and blockage formation, 4) duct wall failure, 5) low-energy disruptive core motion, 6) debris-bed coolability, and 7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt-solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and failure criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the specific phenomena of CDAs in SFRs in details.
Proposal and Demonstration of Strategies for NDT of Conduit Tube of Cable in Conduit Conductors Used in ITER PF Coils
Authors: Chen Zhenmao , Zuo Yongbin , Zhang Dongli , Mao Ying , Huang Taihui
Published: April 2011

The Poloidal Field (PF) coils, made of superconducting Cable in Conduit Conductor (CICC), are important for keeping plasma inside the vacuum vessel of ITER Tokamak device. To ensure the safety of the PF coil, Non-Destructive Testing (NDT) has to be performed for each conduit tube of CICC, as it is the boundary of liquid Helium and supporting structure of the electromagnetic load. However, the complicated cross section shape of the PF conduit tube causes difficulty in its NDT. Aiming at solving the NDT problem of the PF conduit tube, two approaches are proposed and their validity is demonstrated in this paper. One is a hybrid strategy of the ultrasonic phased array and the eddy current array technology. The other is of the immersed ultrasonic testing method. The experimental results reveal that both approaches are promising for the NDT of ITER conduit tubes though the later one is more efficient.
Validation of CATHARE V2.5 Thermal-Hydraulic Code Against Full-Scale PERSEO Tests for Decay Heat Removal in LWRs
Authors: Bandini Giacomino , Meloni Paride , Polidori Massimiliano , Lombardo Calogera
Published: April 2011

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.
Reform of Nuclear Safety Education for Non-Nuclear Engineering Students
Authors: Luo Qinghai , Zou Zhuying
Published: April 2011

The new situations were analyzed for nuclear development and security. Besides nuclear wars and terrorism, reactor runaway and un-ruled radioactive source are the main nuclear accidents in peacetime. With the active development of the nuclear cause in China, the nuclear education system is becoming deficient or defective. In order to ensure the sustainable and efficient accord development of nuclear industry, general nuclear education is necessary for correlative non-nuclear professionals, such as technologies, management, safety, culture and ethic.
Comparison of Selected Forced-Convection Supercritical-Water Heat-Transfer Correlations for Vertical Bare Tubes
Authors: Farah Amjad , King Krysten , Gupta Sahil , Mokry Sarah , Peiman Wargha , Pioro Igor
Published: April 2011

This paper presents an extensive study of heat-transfer correlations applicable to supercritical-water flow in vertical bare tubes. A comprehensive dataset was collected from 33 papers by 27 authors, including more than 125 graphs and wide ranges of parameters. The parameters ranges were as follows: pressures 22.5–34.5 MPa, inlet temperatures 85–350°C, mass fluxes 250–3400 kg/m2s, heat fluxes 75–5,400 kW/m2), tube heated lengths 0.6–27.4 m, and tube inside diameters 2–36 mm. This combined dataset was then investigated and analyzed. Heat Transfer Coefficients (HTCs) and wall temperatures were calculated using various existing correlations and compared to the corresponding experimental results. Three correlations were used in this comparison: Bishop et al., Mokry et al. and modified Swenson et al. The main objective of this study was to select the best supercritical-water bare-tube correlation for HTC calculations in: 1) fuel bundles of SuperCritical Water-cooled Reactors (SCWRs) as a preliminary and conservative approach; 2) heat exchangers in case of indirect-cycle SCW Nuclear Power Plants (NPPs); and 3) heat exchangers in case of hydrogen co-generation at SCW NPPs from SCW side. From the beginning, all these three correlations were compared to the Kirillov et al. vertical bare-tube dataset. However, this dataset has a limited range of operating conditions in terms of a pressure (only one pressure value of 24 MPa) and one inside diameter (only 10 mm). Therefore, these correlations were compared with other datasets, which have a much wider range of operating conditions. The comparison showed that in most cases, the Bishop et al. correlation deviates significantly from the experimental data within the pseudocritical region and actually, underestimates the temperature at most times. On the other hand, the Mokry et al. and modified Swenson et al. correlations showed a relatively better fit within the most operating conditions. In general, the modified Swenson et al. correlation showed slightly better fit with the experimental data than other two correlations.
Optimization Design for a New Cascade of Helium Compressor With Enhanced Pressure Ratio
Authors: Long Yanli , Xu Limin , Yu Jinglei
Published: April 2011

The High Temperature Gas-cooled Reactor (HTGR) is provided with good safety, high quality of thermal source and low cost of power generation in full life cycle. Furthermore, when the helium turbine is used for heat-work conversion, the efficiency of the HTGR is high and up to a magnitude of 50%. One of the key technologies of helium turbine is the helium compressor design. According to the conventional design rule of the air-compressor, the stage number of the helium compressor was too much excessive. Therefore, this thesis has analyzed and optimized a new cascade of helium compressor with enhanced pressure ratio in order to increase the pressure ratio and decrease the stage number. The Artificial Neural Network is used to build the approximate function which is based on database sample space. The Genetic Algorithm is used to search a new design, and the Artificial Neural Network is reused to predict the aerodynamic performance of the new design. The mean camber line and thickness distribution are optimized respectively, and the optimization results show that the total pressure loss coefficient can be reduced by 14.48% than that of the primary.
Three-Dimensional Earthquake Responses of Nuclear Containment Building Installed in Water-Saturated Soil Strata
Authors: Lee Jin Ho , Kim Jae Kwan , Ryu Jeong-Soo
Published: April 2011

In this study, 3-D earthquake responses of nuclear containment building installed in water-saturated soil strata are obtained using a rigorous model of the water-saturated two-phase porous medium. Impedances of a massless rigid foundation of the building and effective earthquake forces are obtained. Far-field region of soil is represented by a 3-D transmitting boundary, which is used for modeling the elastic waves radiating out to an infinite boundary in water-saturated soil strata overlaying rigid bedrock. Using 3-D finite elements for saturated two-phase porous medium, models are developed for the near-field region of soil. The transmitting boundary is combined with the finite-elements model of the near-field region. The impedances and effective earthquake forces are obtained from the numerical model of soil strata. Earthquake responses of the nuclear containment building subject to a 3-D ground motion (two horizontal and a vertical motions) are obtained. The building structures are modeled using a beam-stick model. The impedances of the foundation are combined with the structure model. The effective earthquake forces are applied to the combined model. Floor response spectra of pseudo-acceleration and time histories of acceleration are obtained, which can be used for an analysis of structure-equipment interaction.
Research on the Influence of Ocean Condition on the Oscillation in Parallel Double-Channel
Authors: Zhou Linglan , Zhang Hong
Published: April 2011

In this paper, the flow oscillation in the parallel multichannel system has been studied under ocean condition with RELAP5/MC. A double-channel boiling system model is built using the code RELAP5/MC and it is proved to be capable of simulating the instable phenomena. The influences of rolling, heaving and inclination on the flow oscillation have been analyzed. The result shows that the effect of lengthways rolling on the oscillation in parallel channels can be ignored. The influence of athwartships rolling and heaving reveals resonance effect, that is, when the rolling cycle is close to the oscillation cycle and half cycle respectively, the oscillation would happen earlier, when the difference between the cycles is large, the impact of rolling can also be ignored. The influence of inclination on the oscillation in parallel channels is feeble.
Effect of Temperature in the Preparation of Zirconium Hydride
Authors: Wang Lijun , Zhang Li , Luo Yuanhui , Chen Weidong , Wang Jianwei
Published: April 2011

Zirconium hydride block is an important hydrogen storage material which is used as moderator in space nuclear power reactor. The effect of temperature in the preparation of crackfree zirconium hydride was investigated in this work. The hydriding temperature curve including 3 key steps was designed based on the Zr-H binary phase diagram. The changes of hydrogen pressure with time were observed in the experiment and the hydrogen pressure dropped down when the temperature reached about 876 °C because of the β→δ phase transformation. The phase transformation step in the experiment contributed to the cracking control critically. The hydrogen content of the zirconium hydride was decided by the temperature of the last step.
A Variable Ordering Heuristic Based on Zero-Suppressed Binary Decision Diagrams
Authors: Li Ya-zhou , Wang Jin , Hu Li-qin , Wu Yi-can
Published: April 2011

Two approaches have been proposed to solve the large-scale fault trees or event trees for Probabilistic Safety Assessment in a nuclear power plant. The first one consists in MCS/ZBDD, which uses ZBDDs (Zero-suppressed Binary Decision Diagrams) to implement classical MCS (Minimal Cut Sets) algorithm. The second consists in designing heuristics and strategies to reduce the complexity of the BDDs (Binary Decision Diagrams) construction. This paper was motivated to combine the MCS/ZBDD and designing heuristics for ZBDDs together. A heuristic, which took the failure rate of basic event into account and utilized that truncation could be implemented on ZBDDs during the calculating process, was proposed. This heuristic accelerated the analysis progress by bringing forward the truncation and reducing the complexity of the intermediate ZBDDs. RiskA, a Zero-suppressed Binary Decision Diagram package extended to safety and reliability analysis, has adopted this heuristic. RiskA’s truncation strategies, which had some relations with the ordering scheme, were also introduced. The correctness and efficiency of this new heuristic were verified by some practical models’ analyses.
Mitsubishi Digital I&C System Features for PWR Plant
Authors: Ishimoto Shunsuke , Fumon Hirokazu , Yokoyama Masanori
Published: April 2011

Mitsubishi digital safety Instrumentation and Control (I&C) system has been developed and approved in Japan. The digital I&C system has been applied to many safety and non-safety system applications including full digital I&C system for new plants and digital upgrading for operating plants in Japanese Pressurizer Water Reactor (PWR) plants. The digital I&C system ensures defense-in-depth and diversity for plant safety and control, and this feature also provides countermeasures against software common cause failures. Based on this proven technology, the digital I&C system will also be applied for the US-APWR in the U.S. plant. The US-APWR is one of the candidate reactor of future nuclear power plants in U.S., which has been developed by Mitsubishi Heavy Industries, Ltd. (MHI) modifying Japanese Advanced Pressurizer Water Reactor (APWR) design to comply with U.S codes and standards. The I&C system of the US-APWR also conforms to the U.S. regulatory requirements and industry guidelines. The safety I&C system design and digital platform for the US-APWR are summarized into topical reports and are currently reviewing by U.S. Nuclear Regulatory Commission (NRC). The I&C system includes Human System Interface System (HSIS), Protection and Safety Monitoring System (PSMS), Plant Control and Monitoring System (PCMS) and Diverse Actuation System (DAS). The paper describes our digital I&C design features and application of the digital I&C system to new plants and digital upgrading for operating PWR plants.
Analysis of Operator Responses to Mitigate Temperature Rise in Electrical Auxiliary Building
Authors: Hardy M. James , King Matthew F. , Bigelow Ryan
Published: April 2011

The loss of air conditioning in the electrical auxiliary building at South Texas Project (STP) has been shown in previous work to result in relatively rapid air temperature rise. The heat up is a concern in the Probabilistic Risk Assessment (PRA) because it may be associated with high conditional core damage probability. As a consequence, operator responses to mitigate the building heat up have been developed and proceduralized at STP. The plant’s current loss of electrical auxiliary building HVAC operator procedure was analyzed and improvements have been recommended. This was done using a commercial computational fluid dynamics (CFD) package that modeled the transient air temperature rise of the building and qualitatively assessed the effectiveness of operator actions. A possible methodology for validating the CFD results was developed using coupled energy balance equations for individual rooms. The methodology is a conservative approach to compare the effectiveness operator actions and provide steps for future higher fidelity simulation and validation. While the specific approach is applicable to STP, the overall methods and approaches described should be applicable to other sites that may be subject to excessive room heat up due to loss of air conditioning in critical rooms.
Mission and Status of the First Two Next Generation Nuclear Plant Fuel Irradiation Experiments in the Advanced Test Reactor
Authors: Grover S. Blaine , Petti David A. , Maki John T.
Published: April 2011

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to nine low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and the irradiations will be completed over the next five to six years to support demonstration and qualification of new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of multiple separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) is currently being fabricated and assembled for insertion in the ATR in the early to mid calendar 2010. The design of test trains, the support systems and the fission product monitoring system used to monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the first two experiments will be compared, and updated information on the design and status of AGR-2 is provided. The preliminary irradiation results for the AGR-1 experiment are also presented.
A Stochastic Approach to Nonlinear Seismic Design Spectra
Authors: Brahimi Malek
Published: April 2011

The purpose of this study is to examine the effects of yield strength ratios and damping values on the nonlinear response of Single Degree of Freedom Systems (S.D.F.S) subjected to earthquake ground motion. A stochastic approach to constructing design response spectra and period dependent strength reduction factors for current existing nonlinear design spectra is then proposed. Non-stationary stochastic models are adopted to characterize earthquake ground motion. Twenty simulated earthquakes accelerograms are generated for each of eight historical events using Autoregressive Moving Average (ARMA) techniques. The average of nonlinear response spectra for a given Structural period from a sample of twenty records for each event are calculated to obtain the response spectra. These response spectra are used to examine the effects of structural strength factors such as the yield strength ratio and damping value, and the effects of nonlinear stiffness models including the elastoplasic model, a stiffness degrading model and a stiffness softening model.
Current Status of Conceptual Design Study Toward the Demonstration Reactor of JSFR
Authors: Sakai Takaaki , Kotake Shoji , Aoto Kazumi , Ito Takaya , Kamishima Yoshio , Ohshima Jun
Published: April 2011

JAEA is now conducting “Fast Reactor Cycle Technology Development (FaCT)” project for commercialization before 2050s. A demonstration reactor for Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since FY2007 to determine referential reactor specifications for the next stage of design work of licensing and construction study. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. In this paper, the current status of the conceptual design study for the demonstration reactor plant is summarized.
Experiences From Structural Dynamic Analysis Projects of BWR Plants Within the Scope of Power Uprate Projects Using FEA
Authors: Sva¨rd Bjo¨rn , Larsson Jan-Anders , Ma˚rtensson Philip , Lundin Bjo¨rn
Published: April 2011

During recent years, power-uprate projects have been executed at several BWR-units in Sweden. As part of these projects, structural verification of the safety-related buildings as well as the new and old internal parts of the reactor pressure vessel, RPV, has been performed. In this document, some experiences will be presented from structural dynamic verification, using finite element analysis, FEA, within the scope of these power uprate projects. From this work, a number of conclusions can be drawn. Global models with dense meshes can successfully be used for a broad range of applications. Today, large FEA-models can be used efficiently, e.g. in global vibration and structural verification analyses, if suitable dynamic analysis methods are used. There can be strong dynamic interactions between the containment, fluids, the RPV and RPV-internals. Stress calculation and evaluation can be executed efficiently on large models. The structural models can with advantage be re-utilized in future projects.
Pressurized Water Reactor Core Crud Mapping
Authors: Byers W. A.
Published: April 2011

Knowledge of the quantity and distribution of core deposits (crud) can be useful in many ways. Trending the amount of crud can show the effects of coolant chemistry changes and crud remediation efforts. The crud distribution reflects the locations of sub-cooled boiling, and indirectly, core performance. Both the quantity and distribution of crud can be useful in assessing the risk of Crud Induced Power Shifts (CIPS) and Crud Induced Local Corrosion (CILC). Measuring core crud is also essential in increasing the fundamental understanding of crud transport and the build-up of ex-core radiation fields. Several techniques have been used to measure core crud levels. Crud scraping is frequently applied to sample crud, but it cannot be applied economically over broad areas. Crud has been mapped over broad core areas by analyzing the waste stream from an ultrasonic fuel cleaner, but uncertainties in the cleaning efficiency limit the accuracy of this technique. Video crud mapping, the determination of crud coverage from video images of fuel assemblies, is an economical method that can be readily applied to peripheral rods on every fuel assembly. The mapping can be performed by computer image analysis, or by visual comparisons to standards. The two techniques are compared and the advantages and disadvantages of each are discussed. The effectiveness of video crud mapping in assessing fuel performance issues is shown in the results of several field applications.
Investigation on Integrity of JMTR Reactor Pressure Vessel
Authors: Ide Hiroshi , Kimura Akihiro , Miura Hiroshi , Nagao Yoshiharu , Hori Naohiko , Kaminaga Masanori
Published: April 2011

Visual observation of inner side of a reactor pressure vessel of Japan Materials Testing Reactor (JMTR) was carried out using an underwater camera before the JMTR refurbishment work from the view point of its long term utilization, because the reactor pressure vessel of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed using the gauze were observed around nozzles in a top closure of the reactor pressure vessel. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the reactor pressure vessel was observed, and then the integrity of the reactor pressure vessel was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the reactor pressure vessel of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the reactor pressure vessel in future.
Influence of an Impinging Jet on Sedimented Debris
Authors: Kratzsch Alexander , Renger Stefan , Ka¨stner Wolfgang , Hampel Rainer , Krepper Eckhard
Published: April 2011

The presented work is part of a joint research project performed in cooperation between the Forschungszentrum Dresden and University of Applied Sciences Zittau/Goerlitz. The paper deals with experimental investigations concerning the influence of an impinging jet on sedimented insulation material in the building sump of the reactor containment. One of the main tasks in reactor safety research is the safe heat dissipation from the reactor core and the containment of light-water reactors. In the case of loss of coolant accident (LOCA) the possibility of the entry of insulation material into the containment and the building sump of the containment and into the associated systems to the residual heat exhaust is a serious problem. In the long-term phase of a LOCA the coolant ejected by the pipe leakage falls several meters onto the sump water surface. On this way, the jet is mixed with air. Furthermore, the impinging jet will entrain air bubbles into the building sump of the reactor containment. The entrained air bubbles will rise and have an additional influence on the flow field in the sump and of the sedimented insulation material. Hence, the impinging jet has an influence of the insulation material transport in the sump. To investigate the influence of an impinging jet a special test facility was designed. The test facility “Tank” was build up with acrylic glass. Thereby it is possible to use laser PIV to measure the flow field and high-speed video to analyze the jet-structure in the test facility. With help of this instrumentation the following experiments were performed: • Experiments without air entrainment and different distances between pipe outlet and water surface in the test facility. • Experiments with air entrainment, different distances between pipe outlet and water surface in the test facility and different flow velocities of the impinging jet. • Experiments with a cold impinging jet and hot water in the test facility for a defined distance between pipe outlet and water surface and a defined jet velocity of the jet. The main goal of the experiments is to study the physical phenomena of the impinging jet and provide experimental data for verification of CFD models. The corresponding CFD investigations are reported by Krepper et al. [KRE10].
Integrating Cyber Security Into Nuclear Digital I&C Safety Systems
Authors: Zhang Deanna Jing
Published: April 2011

Due to the digital modernization of nuclear power plants, along with the prevalence of connectivity between safety systems and other systems, cyber security has become a significant issue during the development and operation of nuclear digital instrumentation and controls (I&C) safety systems. To mitigate this issue, cyber security may be integrated into the design and development process of digital safety systems. Cyber security features that ensure confidentiality, integrity, and availability may be integrated into the design of these systems. In addition, a robust information security program should be implemented to secure the development environment of these digital systems to prevent malicious manipulation of the system while under development. However, concerns have been raised regarding the effect that such cyber security features may have on the performance and operation of safety systems. The intent of this paper is to address these concerns by demonstrating how cyber security features can be integrated into the design of safety systems. Specifically, this paper demonstrates how several safety features can also be utilized to enforce cyber security, and how several additional cyber security features may be integrated into the safety system design without degrading safety functions. Furthermore, this paper discusses how configuration management and access control during the development process can mitigate cyber security concerns by preventing unauthorized manipulation of the system while under development.
Research on Distribute Real-Time Database Based on Vxworks
Authors: Li Huwei , Li Duo , Xiong Huasheng
Published: April 2011

Distributed control system (DCS) is widely used in industrial process control. As a matter of fact, the database in DCS is of great importance for the safety and reliability of the whole system. This paper focuses on the distributed real-time database in a dedicated DCS used to a small nuclear power station. A database which runs under Vxworks embedded operation system and can satisfy the basic database requirements was established by using the RDM Embedded database management system which reduces the operation system resources consumption. Three redundant databases are running in three distributed embedded computers, and the UDP protocol is used to implement their synchronization and the data transmission from low level computers to them. This system architecture is able to enhance the database reliability and the overall system performance. Experimental testing results show that the developed redundant databases structure running in three embedded computer called operator stations is efficient and reliable to manage real-time data, and it can meet its function and performance requirements in small nuclear power plant.
Polarized by the Light
Authors: Wong Wei
Published: April 2011

Light-dependent phosphorylation of the auxin efflux carrier PIN3 determines its trafficking pattern in the hypocotyl endodermis.
Energy Requirements for Neuronal Polarization
Authors: Hurtley Stella M.
Published: April 2011

A bioenergy-sensing pathway determines axon initiation and growth in neurons.
Major Basic Protein from Eosinophils and Myeloperoxidase from Neutrophils Are Required for Protective Immunity to Strongyloides stercoralis in Mice
Authors: O'Connell Amy E. , Hess Jessica A. , Santiago Gilberto A. , Nolan Thomas J. , Lok James B. , Lee James J. , Abraham David
Published: April 2011

Eosinophils and neutrophils contribute to larval killing during the primary immune response, and neutrophils are effector cells in the secondary response toStrongyloides stercoralisin mice. The objective of this study was to determine the molecular mechanisms used by eosinophils and neutrophils to control infections withS. stercoralis. Using mice deficient in the eosinophil granule products major basic protein (MBP) and eosinophil peroxidase (EPO), it was determined that eosinophils kill the larvae through an MBP-dependent mechanism in the primary immune response if other effector cells are absent. InfectingPHILmice, which are eosinophil deficient, withS. stercoralisresulted in development of primary and secondary immune responses that were similar to those of wild-type mice, suggesting that eosinophils are not an absolute requirement for larval killing or development of secondary immunity. TreatingPHILmice with a neutrophil-depleting antibody resulted in a significant impairment in larval killing. Naïve and immunized mice with neutrophils deficient in myeloperoxidase (MPO) infected withS. stercoralishad significantly decreased larval killing. It was concluded that there is redundancy in the primary immune response, with eosinophils killing the larvae through an MBP-dependent mechanism and neutrophils killing the worms through an MPO-dependent mechanism. Eosinophils are not required for the development or function of secondary immunity, but MPO from neutrophils is required for protective secondary immunity.
Role of Gastric Epithelial Cell-Derived Transforming Growth Factor β in Reduced CD4
+
T Cell Proliferation and Development of Regulatory T Cells during Helicobacter pylori Infection
Authors: Beswick Ellen J. , Pinchuk Iryna V. , Earley Rachel B. , Schmitt David A. , Reyes Victor E.
Published: April 2011

Gastric epithelial cells (GECs) express the class II major histocompatibility complex (MHC) and costimulatory molecules, enabling them to act as antigen-presenting cells (APCs) and affect local T cell responses. During
Helicobacter pylori
infection, GECs respond by releasing proinflammatory cytokines and by increasing the surface expression of immunologically relevant receptors, including class II MHC. The CD4
+
T cell response during
H. pylori
infection is skewed toward a Th1 response, but these cells remain hyporesponsive. Activated T cells show decreased proliferation during
H. pylori
infection, and CD4
+
CD25
+
FoxP3
+
regulatory T cells (Tregs) are present at the site of infection. In this study, we examined the mechanisms surrounding the CD4
+
T cell responses during
H. pylori
infection and found that transforming growth factor β (TGF-β) plays a major role in these responses. GECs produced TGF-β1 and TGF-β2 in response to infection. Activated CD4
+
T cells in culture with
H. pylori
-treated GECs were decreased in proliferation but increased upon neutralization of TGF-β. Naïve CD4
+
T cell development into Tregs was also enhanced in the presence of GEC-derived TGF-β. Herein, we demonstrate a role for GEC-produced TGF-β in the inhibition of CD4
+
T cell responses seen during
H. pylori
infection.
Structural Characterization of the Erythrocyte Binding Domain of the Reticulocyte Binding Protein Homologue Family of Plasmodium yoelii
Authors: Grüber Ardina , Gunalan Karthigayan , Ramalingam Jeya Kumar , Manimekalai Malathy S. S. , Grüber Gerhard , Preiser Peter R.
Published: April 2011

Invasion of the host cell by the malaria parasite is a key step for parasite survival and the only stage of its life cycle where the parasite is extracellular, and it is therefore a target for an antimalaria intervention strategy. Multiple members of the reticulocyte binding protein homologues (RH) family are found in all plasmodia and have been shown to bind to host red blood cells directly. In the study described here, we delineated the erythrocyte binding domain (EBD) of one member of the RH family, termed Py235, from
Plasmodium yoelii
. Moreover, we have obtained the low-resolution structure of the EBD using small-angle X-ray scattering. Comparison of the EDB structure to other characterized
Plasmodium
receptor binding domains suggests that there may be an overall structural conservation. These findings may help in developing new approaches to target receptor ligand interactions mediated by parasite proteins.
Traffic risks by travel mode in the metropolitan regions of Stockholm and San Francisco: a comparison of safety indicators
Authors: McAndrews Carolyn
Published: April 2011

According to commonly used measures of traffic safety, Sweden has one of the safest road transportation systems in the world, whereas the USA has relatively poor road safety performance. Although national comparisons are useful, they are problematical because they generalise across a diverse mix of travel environments (eg, urban and rural). This study used an array of traffic death rates to determine whether comparable urban regions in Sweden and California—Stockholm and San Francisco—have similar road safety performance for various types of road users. The study found that the Stockholm region is far safer than the San Francisco Bay area for pedestrians and bicyclists, even when comparing the regions' core cities, but may not be any safer for motor vehicle occupants. In addition, comparing traffic safety with traditional measures of exposure such as population and motor vehicle travel produced different results than measures that account for mode-specific exposure.