bluuebunny/crossref_metadata_2025 huggingface dataset:
Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3
Authors: Akers Douglas W. , Harvego Edwin A.
Published: October 2013

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1–3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI-2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the reactor pressure vessel will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues are included in the paper.
A Study on Phase Difference Between Pressure Drop and Flow Rate of Sine Fluctuation in Rectangular Channel
Authors: Zhou Bao , Gao Pu-zhen , Tan Si-chao , Tian Jing-da
Published: October 2013

An experimental investigation on fluctuating turbulent flow with different amplitudes, frequencies and mean values of flow rate in a narrow rectangular channel was carried out to determine the phase difference so as to find out real corresponding relationship between pressure drop and flow rate. It is found that the measurement delay time difference between the flow meter and the differential pressure transmitter is not a constant but vary with the different flow rate fluctuation conditions. The phase difference was calculated by a function which is given in this paper and tested by the result of two kinds of nonlinear fit methods, whose results agree well.
Dynamic Analysis of S-CO2 Cycle Control With Coupled PDC-SAS4A/SASSYS-1 Codes
Authors: Moisseytsev Anton , Sienicki James J.
Published: October 2013

The ANL Plant Dynamics Code (PDC) for the analysis of supercritical carbon dioxide (S-CO2) Brayton cycle power converters has been under development at Argonne National Laboratory for several years. It is the leading state of the art capability for modeling S-CO2 cycles at the system level. The SAS4A/SASSYS-1 Liquid Metal Reactor Code System is the leading capability for modeling a SFR or LFR at the system level. The SAS4A/SASSYS-1 code combines reactor dynamics with thermal hydraulics calculations. Coupled together, the two codes allow users to carry out for the first time very detailed transient calculations of an entire plant incorporating a SFR with a S-CO2 cycle energy converter, including simulation of both design and beyond-design events. The developed coupling approach utilizes the restart capability of SAS4A/SASSYS-1 and allows the data transfer between the two codes at each time step. The coupling approach is demonstrated by carrying out S-CO2 cycle control calculations in which the entire plant follows a user-defined electrical grid demand scenario. The automatic control mechanisms on the S-CO2 cycle side allow following of the load changes in the entire 0–100% range. On the reactor side, an autonomous (i.e., no deliberate control other than startup and shutdown) operation is assumed where the reactor power is adjusted by the internal reactivity feedbacks to match the heat removal rate by the balance-of-plant.
Modeling of the SNL S-CO2 Loop With ANL Plant Dynamics Code
Authors: Moisseytsev Anton , Sienicki James J.
Published: October 2013

The ANL Plant Dynamics Code (PDC) for the analysis of supercritical carbon dioxide (S-CO2) Brayton cycle power converters has been under development at Argonne National Laboratory for several years. In previous years, limited validation of the PDC models on an individual basis was carried out using experimental data obtained from facilities directed at individual components. Recently, experimental data from the SNL/BNI small-scale S-CO2 Brayton cycle demonstration that is being assembled in a staged fashion has been provided to ANL. The loop configuration with a single turbo-alternator-compressor (TAC) was modeled with the Plant Dynamics Code and the performance prediction of individual components and the entire loop obtained from the PDC was compared with the experimental data. Overall, reasonably good agreement is obtained for steady state conditions around the loop when a value is inferred for the heat loss downstream of the turbine such that a steady state loop energy balance can be established.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration — (5) Measurement of Bubble Deformation Near Wall Under Structure Vibration
Authors: Mizuno Kousuke , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki
Published: October 2013

In a nuclear power plant, one of the important issues is evaluation of the safety of reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs. However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there are little knowledge how coolant in a core response when large earthquake acceleration is added.

Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction or the power of core is fluctuated with the oscillation by the experiments and numerical analysis. However detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow structure. In order to investigate it, we visualize changing of bubbly flow structure in pipeline on which sine wave is applied.

Bubbly flow is produced with injecting gas into liquid flow through a horizontally circular pipe. In order to vibrate the test section, the oscillating table is used. The frequency of vibration added from the table is from 1.0 Hz to 10 Hz and acceleration is from 0.4 G to 1 G (1 G = 9.8 m/s2). The test section and a high speed video camera are fixed on the table. Thus the relative velocity between the camera and the test section is ignored.

In the visualization experiment, the PIV measurement is conducted. Then the motion of bubbles, for example the shape, the positions and the velocity are measured with observation.

In addition, by varying added oscillation amplitude, frequency and flow rate of the fluids, the correlation between these parameters and bubble motion was evaluated.

It was clarified that the behavior of liquid phase and bubble through horizontal circular pipes was affected by an oscillation. When structure vibration affects the flow, two main mechanisms are supposed. One is the addition of body force of the oscillation acceleration to liquid phase and bubble, and the other is the velocity oscillation of the test section and the effect of the boundary layer of the pipe wall. It was also found that when the added oscillation frequency and amplitude was changed, the degree of the fluctuation of liquid phase and bubble motions were changed.
Slug Flow Main Parameters in a Combined Inclined Up and Down Two-Phase Flow
Authors: Sanchez-Silva Florencio , Carvajal-Mariscal Ignacio , Toledo-Velázquez Miguel , Polupan Georgiy
Published: October 2013

Slug flow induced by the combination of a downstream and an upstream two phase flow was experimentally studiedin this paper. A loop, where it is possible to change the flow conditions and the lengths and inclination angles of the pipe sections, was built. Three conductive probes were used for the hold-up measurement in different sections. A data acquisition system based on a microcontroller was also developed. The experimental runs were set for four alternated upstream and downstream sections. Several gas and liquid flow rates to get stratified flow in the first section, were also tested in order to observe the hold-up variation of the resulting slugs. Barnea’s criterion predicts acceptably the severe slugging trend for the tested experimental conditions, but for stable slug flow, it gives an incorrect idea of the phenomenon because it doesn’t account the void fraction in the column, which produces a value hL,min < 1, and there is still liquid flowing into the separator. On the other hand, the method doesn’t consider the existence of a pseudo slug regime in which there is a combination of stable slug flow and sporadic formation of severe slugging. Due to this situation, the flow pattern depends greatly on the amount of liquid draining back that increases the liquid accumulation rate and when the liquid attains its critical level a sudden slug takes place. To improve the method proposed by Barnea, the void fraction has been taken into account. According to the experimental results, there is an important influence of the geometry, that is the length and inclination angles of the sections, in the magnitude of the main slug flow parameters.
Development and Implementation of a 3D Heat Conduction Model for (Very) High Temperature Reactors Into the Reactor Dynamics Code DYN3D
Authors: Baier Silvio , Rohde Ulrich , Kliem Soeren , Fridman Emil
Published: October 2013

The reactor dynamics code DYN3D was extended to treat phenomena in Block-type High Temperature Reactors (HTR). Therefor, a new heat conduction model was implemented into the code to tackle 3D effects of heat conduction and heat transfer. The first part of the paper describes the details of the heat conduction model. In the second part results of coupled neutron-kinetics/thermal-hydraulics calculations of steady state and short-time transients in block-type HTRs are discussed.
Numerical Simulation of the Self-Leveling Phenomenon by Modified SIMMER-III
Authors: Zhang Bin , Matsumoto Tatsuya , Morita Koji , Yamano Hidemasa , Tagami Hirotaka , Suzuki Tohru , Tobita Yoshiharu
Published: October 2013

During a hypothetical core-disruptive accident in a sodium-cooled FBR, degraded core material can form debris beds on the core-support structure and/or in the lower inlet plenum of the reactor vessel, due to the rapid quenching and fragmentation of the core material melt. Heat convection and vaporization of the sodium will lead ultimately to leveling the debris bed that is of crucial importance to the relocation of the molten core, the recriticality evaluation and the heat removal capability of the debris bed. There is, therefore, a great need for more studies focusing on this topic, especially the much needed numerical simulation. The widely-used fast reactor safety analysis code, SIMMER-III, has difficulties in this simulation because of the lack of modeling for mechanistic interactions among particles in the current version. However, the extensive experimental analysis and the previously-proposed analytical model provide SIMMER-III the possibility of taking consideration of the extra influence of solid particles in this phenomenon. Thus, the debris fluidization model and the boiling regulation model are proposed and introduced into SIMMER-III. Calculations, by the modified SIMMER-III, against several representative experiments with typical self-leveling behavior have been performed and compared with the evaluated items recorded in experiments. The good agreements on these items suggest the modified SIMMER-III can simulate the self-leveling behavior with reasonable precision, especially on the onset of self-leveling, although further model improvement is necessary to represent the transient behavior of bed leveling more reasonably.
Current Technology for Power Plant Makeup Water Treatment and Wastewater Recovery
Authors: Buecker Brad
Published: October 2013

The availability of fresh water for power plant makeup and cooling continues to decline. In addition, environmental regulators are placing tighter limitations on the methods by which water is obtained and discharged for and from industrial processes. One example is the emerging restrictions on once-through cooling in the U.S. This paper examines state-of-the art technology for producing plant makeup water, and it also looks at techniques for recovering significant portions of wastewater streams such as cooling tower blowdown. Technologies that will be reviewed include makeup water pretreatment, reverse osmosis, ion exchange, electrodeionization, and high-pH reverse osmosis for wastewater recovery.
Molecular Dynamics Simulations to Evaluate the Effect of Applied Strain on Interstitial Cluster Formation and Orientation Under Collision Cascade Damage
Authors: Miyashiro Satoshi , Fujita Satoshi , Itakura Mitsuhiro , Okita Taira
Published: October 2013

We conducted molecular dynamics (MD) simulations to analyze the strain influence on defect formation and orientation. Collision cascade damage was initiated under uniaxial applied strain with a PKA energy of 10 keV. The number of residual defects increased with applied strain because of the enhanced formation of larger defect cluster. We also applied uniaxial strain to the simulation cell which included an interstitial cluster and detected the change in its direction. The probability of a change in the defect cluster direction was significantly higher under strain. Results further showed that the probability of the change in direction is higher with smaller defect clusters, and that it is extremely low with clusters larger than a certain size.
Sensitivity Analysis of CFD Code FLUENT-12 for Supercritical Water Heat Transfer in Vertical Bare Tubes
Authors: Farah Amjad , Haines Patrick , Harvel Glenn , Pioro Igor
Published: October 2013

The ability to use FLUENT 12 or other CFD software to accurately model supercritical water flow through various geometries in diabatic conditions is integral to research involving coal-fired power plants as well as Supercritical Water-cooled Reactors (SCWR). The cost and risk associated with constructing supercritical water test loops are far too great to use in a university setting. Previous work has shown that FLUENT 12, specifically realizable k-ε model, can reasonably predict the bulk and wall temperature distributions of externally heated vertical bare tubes for cases with relatively low heat and mass fluxes. However, sizeable errors were observed for other cases, often those which involved large heat fluxes that produce deteriorated heat transfer (DHT) regimes, which involves a reduction in the rate of heat transfer between the wall and the fluid, and therefore a rise in wall temperature. These errors were believed to be caused by FLUENT’s over-estimation of changes in thermal physical properties of the fluid which occur around the pseudocritical point.

The goal of this research is to gain a more complete understanding of how FLUENT 12 models supercritical water cases and where errors can be expected to occur. One control case is selected where expected changes in bulk and wall temperatures occur and they match empirical correlations’ predictions, and the operating parameters are varied individually to gauge their effect on FLUENT’s solution. The model used is the realizable k-ε, and the parameters altered are inlet pressure, mass flux, heat flux, and inlet temperature.

As a result of this work, under supercritical conditions, the only finite limitations found in the k-ε model are in the mass flux which is predicted correctly above 300 kg/m2s, and in the temperature and related physical properties of water where they exceed the limitation of the fluid properties in the fluid database.
Risk-Informed and Defense-in-Depth Oriented Plant Design Approach
Authors: Morota Hidetsugu , Suzuki Hiroshi , Sakihama Takashi , Nasif Hesham R. , Sano Hiroshi
Published: October 2013

Defense-in-Depth is the basis of safety design of nuclear plants, and refined and strengthened year by year. Nowadays, the importance of it has been further highlighted triggered by the accident at Fukushima Daiichi Nuclear Power Station.

Accidents described previously have been shown it could not care enough to uncertainty related to design, construction, maintenance and operation has exposed. It’s the lessons learned of that is to say to reduce uncertainty of design is the use of risk informed is essential. Therefore, the establishment of the design approach that uses risk informed consideration of the Defense-in-Depth is an important theme.

Defense-in-Depth is a measure which prevents the increase of the event frequency and core damage in consideration of the degree of safety margins, redundancy, diversity and consideration of radiation safety due to core damage and security.

The plant designers and utilities have made efforts to ensure safety utilized conventional design technique, which means deterministic, and risk information, in order to incorporate Defense-in-Depth concepts.

To consider Defense-in-Depth in the design phase, various requirements should be taken into consideration. It has been coming to be able to perform more rational and quantitative judgment by utilization of risk information.

In that case, while the designers of various fields will work in cooperation to ensure safety, if there are common utilization schemes for risk information among designers, more efficient and rational design works can be advanced in consideration of Defense-in-Depth. However, conventionally, it is hard to say that such schemes have functioned in order that designers may advance design works in collaboration.

This research intends to generate the schemes which advance design works in sharing the same risk information databases, which mean various risk indications etc), in other words, the schemes which will become lingua franca for utilizing risk information, in case of nuclear plants designs.
Teaching of Nuclear Power Plant Safety at the National Technical University of Ukraine Within the New Course in Safety Culture
Authors: Begun Vasilij V. , Begun Sergij V. , Kilina Olena O.
Published: October 2013

The necessity of safety analysis methods and probable scenarios of accidents teaching in the education of experts for nuclear industry in Ukraine has been realised only after the Chernobyl accident. We developed the content of the first educational course in probabilistic safety analysis in 1995 based on the experience of the countries having developed nuclear power, the USA first of all, and on the training course of the Idaho National Laboratory. After this in 1996 the new course in probabilistic safety analysis of nuclear power plant (NPP) was adopted at our university.

The new educational course in safety for students was developed and adopted in 2009 educational year - “Safety culture at nuclear installations of Ukraine”. Education and training in safety culture in higher educational institutions and in the nuclear power plants is a part of the general modern process of maintenance of safety, it is recommended by IAEA standards. The principles of safety culture are taken as a basis of the modern concept of safety of nuclear power plants.

This work has received a positive appreciation from the management of departments of safety and training of the personnel of operating organization National Nuclear Energy Generating Company Energoatom (NNEGC Energoatom) and from other leaders of nuclear industry. The content of this educational course was discussed at the international scientific conferences on safety culture in 2008 and 2010, and was preliminary printed in the professional journal «Nuclear and radiation safety». The purposes of education have been defined as a survey, generalizing course on safety of the NPP with an allocation of safety issues on the foreground. Practical questions of the equipment and NPP systems work, their interaction in emergencies and the role of the human-operator are studied. The procedure of failure analysis at NPP is studied. Students analyze equipment work, root and direct causes of incidents. Methods of estimation of safety conditions based on observable operational indicators are studied. Parameters, variables and indicators of safety culture are studied.

As a result of gained experience we have come to the conclusion about high advisability of educational courses in safety for students. Specially formed knowledge and education in the field of safety from a student’s bench are the basis of safety culture of the future nuclear industry expert.
Module Fabrication Strategy for Today’s Nuclear Industry
Authors: Smith Clayton T. , Hammeran John H. , Lockwood Carl
Published: October 2013

First generation nuclear power plants were built onsite with large construction forces over a number of years using a design that was being revised during construction. Vendor supplied equipment skids were the closest thing to modular building techniques at that time. Modularization techniques are available today as a tool for performing parallel construction activities such as fabrication and assembly of components offsite to support the construction cost and schedule goals of a project. Accordingly, extensive planning and coordination is required by engineering, procurement, fabricators, and construction to support modularization.

The importance of developing a strategy for the utilization of modularization to the Nuclear Industry, Owners, and Engineering Procurement and Construction (EPC) entities, cannot be over emphasized, ensuring the development of a Corporate Modularization Fabrication Strategy, which incorporates and addresses the following, various elements:

• Project Modularization Plan

• Definitions and terms for module, modularization, and on-off-site fabrication

• Module types • Modularization boundaries

• Work processes for interface between Engineering, Procurement, Fabricator, and Quality to allow good communication between entities

• Constructability reviews • Modularization schedule linked to integrated project schedule

• Compliance with approved procedures and Quality Program requirements

• Commitment from management

This paper and presentation will discuss and highlight the following for a typical Advanced Boiling Water Reactor (ABWR) application:

• Typical Corporate Modularization Fabrication Strategy;

• Typical Modularization Concepts;

• Modularization: Friend or Foe.
Operating Condition of Steam Injector as a Passive Cooling System at Severe Accident of Nuclear Power Plant
Authors: Abe Yutaka , Shibayama Shunsuke , Kaneko Akiko , Iwaki Chikako , Narabayashi Tadashi , Mori Michitsugu
Published: October 2013

Steam injector (SI) is a passive jet pump which is driven by high-performance steam condensation onto water jet and it is expected to be active at severe accident of nuclear power plant with no electricity. SI is mainly consists of convergent-divergent nozzle. Supersonic steam flow condenses onto water jet in the mixing nozzle and mass, momentum, and energy of steam is transferred to water in the mixing nozzle. Condensed water jet is accelerated at the throat and kinetic energy is converted into pressure in the diffuser, which produces higher pressure than inlet steam pressure. It is easy to apply the SI to nuclear power plant since SI has quite simple and compact structures.

The objectives of the present study are to clarify the mechanism of heat and momentum transfer in the mixing nozzle and to determine operating range of SI for practical use.

A transparent test section is adopted to conduct visualization of the flow structure with a high-speed video camera as well as measurement of pressure distribution in mixing nozzle, throat, and diffuser with changing back pressure. Fundamental parameters change between operative and inoperative state of the injector were evaluated by measuring pressure and temperature distribution along axial direction of the test section. Discharge pressure as one of operating characteristics of the injector was also measured in changing back pressure by decreasing the opening ratio of the back pressure valve attached downstream of the test section.

It was confirmed that discharge pressure increased and the injector became inoperative unsteadily with decreasing opening ratio of the back pressure valve just after it produced the maximum discharge pressure. In the present investigation, this maximum discharge pressure is evaluated as the operation limit of the injector. Furthermore, discharge pressure from diffuser, which is one of the indicators of operating performance as well as operating limit is predicted from inlet condition adopting one-dimensional analysis model proposed previously. By comparing analytical result with experimental data, as well as visualization of flow structure in throat and diffuser, physics model including two-phase flow structure with shock wave which was observed at throat and diffuser are discussed in order to predict injector’s operation with high accuracy.
A New Approach to Calculating Fuel Temperature in TRIGA Mark I Research Reactors
Authors: Johns Jesse M. , Reece W. D.
Published: October 2013

Often a forced convection heat transfer coefficient is used to calculate the peak fuel temperature for a rectangular lattice TRIGA core even though the core is cooled by natural convection. The arguments for applying a forced convection empirical relationship are examined and another relationship is suggested. The peak fuel temperature was calculated using two different correlations, Dittus-Boelter and natural convection, for pool temperatures of 30°C and 60°C. The Dittus-Boelter correlation predicted a fuel temperature rise of 1.85°C for this difference in pool temperature, contrary to the predicted rise of 25.64°C from natural convection relationships. Experimental data shows that the relationship of fuel temperature rise with increasing pool temperature is more accurately represented by the natural convection correlation than with Dittus-Boelter. Using a derived natural convection correlation, the calculated peak fuel temperatures then closely match measured data. A procedure was developed to access convective heat transfer coefficient changes in the cladding gap as a function of reactor power for the hot channel which are similar to those presented in literature.
Visual Experiments on the Subcooled Water Droplet Behavior in High Temperature Molten Metal
Authors: Zhou Yuan , Lin Meng , Zhang Zhengming , Lei Lei
Published: October 2013

Water injection mode of molten fuel/coolant interaction is a key issue during the steam generator tube rupture in liquid metal reactors. The fragmentation behaviors of plunging water droplet into a high temperature molten tin liquid pool were studied by using a high speed video camera. Photographs of the water droplet configuration, the vaporization at the interface between the melt and water, and the vapor film disturbance were obtained. The results demonstrated that the water droplet in the molten tin pool was near spherical shape initially and the vapor was generated and accumulated in the bottom of the water droplet. The steam explosions happened when vapor film collapses finally.
Generation and Mechanical Analysis of Repeatable Random Rough Surfaces
Authors: Medina Hector E. , Hinderliter Brian
Published: October 2013

Natural weathering and wear of structural materials in service nearly always generate surface roughness, and follow the Central Limit Theorem prediction for surface topology. This study couples experimental and statistical theory, to extend knowledge of life of materials from initial service surface conditions through surface damage accumulation. Statistical moments and other parameters were correlated with fracture locations probability (H/N), versus auto correlation length, and depth. As the surface grows to a full Gaussian, H/N increases its dependence on profile’s average and RMS roughness, and derived parameters. This dependence shows an asymptotic limit behavior that approaches agreeably Griffith’s crack criterion, though with multiple fracture locations. Importantly, a Transitional Region was observed, below which the probable location of fracture is uncorrelated to the parameters studied. This is because introduced roughness is insufficient to compete with impurities, internal and external manufacturing flaws, and scratches, due to handling and machining, on the samples.
Effect of CRGT Cooling on Modes of Global Vessel Failure of a BWR Lower Head
Authors: Villanueva Walter , Tran Chi-Thanh , Kudinov Pavel
Published: October 2013

An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head.

A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis.

It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) ‘ballooning’ of the vessel bottom for smaller pools, and (b) ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.
Observations of Irradiation-Induced Defect Formation on Si (111) Surface by a Linked Facility Between an Ion Accelerator and STM
Authors: Sato Motohiro , Kikuchi Tairo , Murakami Kenta , Miyashiro Satoshi , Okita Taira
Published: October 2013

A linked facility between an ion accelerator and STM enabled to elucidate cascade damage process experimentally with an atomistic scale.

By adjusting experimental conditions, we have successfully attained very clear images of Si (111) surface irradiated by very low-flux and stable ion beams. We can obtain the information of displacement damage produced by one injected ion. Much higher density of defects is observed than the number of injected ions throughout the area of the image, indicating that there are processes for defect formation other than direct collision with injected particles.
Visualization and Numerical Simulation of Exchange Flow in Density Different Gases
Authors: Fumizawa Motoo , Ohkawa Shuhei , Buma Isaku , Tanaka Suguru
Published: October 2013

Buoyancy-driven exchange flows of helium-air through inclined a narrow tube was investigated. Exchange flows may occur following the opening of a window for ventilation, as well as when a pipe ruptures in a high temperature gas-cooled reactor. The experiment in this paper was carried out in a test chamber filled with helium and the flow was visualized using the smoke wire method. A high-speed camera recorded the flow behavior. The image of the flow was transferred to digital data, and the slow flow velocity, i.e. micro flow rate was measured by PIV software. Numerical simulation was carried out by the code of moving particle method with Lagrange method.
How Vertical Project Team Integration Between End User, End User’s Engineer, Nuclear QA, and Equipment Vendor Provided for Minimized Cost and Installation Expense
Authors: Tyman Jordan , Bruno Joe , Holzman Perry
Published: October 2013

In today’s economic environment, larger capital investments by power plants, especially nuclear plants, are tending towards shorter project lead times; and at the same time with an emphasis on improved performance and future reliability. These ideas are generally not synonymous with cost effectiveness. Often it is very beneficial that the engineering phase for this equipment be flexible such that the needs of the end user, engineer, and equipment vendor are met in a cooperative fashion.

The recent power uprate scope of supply at NextEra Energies Point Beach Nuclear Power Plant was modified about 36 weeks before the outage start date to include large replacement condensate coolers for each unit with the Unit 2 exchanger delivery being critical. This left a small window in which to generate a specification, purchase, and produce the first of two 14,500 ft2 heat exchangers. The project had many design challenges that needed to be addressed including the installation, size limitations, nozzle locations, nuclear plant quality requirements and consideration of lake/sea grass (string from algae) fouling, zebra/quagga mussel fouling, and alewife (small fish) pluggage concerns in order to bring the project to a successful conclusion. These issues needed the combined strengths of the Engineer, End User, and Fabricator working quickly and making timely decisions to bring the total project to the most cost effective and successful conclusion.
Research on Nuclear Heat Deposition Behavior in the Spallation Target of an Accelerator Driven Subcritical System
Authors: Xiao S. , Yang Y. , Zhou Z.
Published: October 2013

The power density distribution behavior of the ADS spallation target, which is a key factor in the thermal-hydraulic and mechanical design of the high-power ADS target, was investigated under different proton incident energy. A Chinese ADS conceptual design of spallation target was proposed in this paper. The deposition heat in the spallation target was calculated by MCNPX code. From the results, it was found that the Bragg peak phenomenon weakens as the proton source incident energy increases. Large Bragg peak was observed for proton incident energy below 500MeV, however for the proton source energy above 900MeV, Bragg peak phenomena was not obvious. Analysis on the nuclear reactions behavior and ionization process induced by source proton in the target was carried out to address this issue. Meanwhile, the results show that the proton leakage rate from the target, which is an important factor in proton radiation shielding, greatly depends on the proton range (penetration depth) under different incident energy. In order to stop or keep the protons in the target, the minimum axial length (in the incoming particles direction) of the target at a given proton source incident energy should be determined according to the corresponding proton range. The results of this paper will be useful to guide the design of spallation target of a reference ADS.
Main Generator Life Cycle Management
Authors: Puri Arun , DiBiase John
Published: October 2013

The Ginna Nuclear Power Plant began commercial operation in 1970. Two power uprates have been completed during the 40+ years of operation and the original main generator continues to provide highly reliable operation with no significant changes to the original design. The life cycle management plan for the generator identified several age-related issues which could challenge reliable operation of the unit if left unattended. This paper outlines the approach taken to assess and quantify specific component degradation, the impact on generator reliability and the counter measures (planned and implemented) to ensure continued reliable operation until a generator replacement is performed in a future refueling outage.
Tube Pullout Testing Experience
Authors: Massey Kim M. , Jones Merwin W.
Published: October 2013

This paper provides an overview of the use of pullout testing to determine the optimum parameters for assuring reliable tube to tubesheet joints in condensers and heat exchangers. The test is described with typical test methods, equipment and test precautions. Numerous examples of data from a variety of units, tube materials, and tubesheet materials are presented. Also included are unique tests which were performed to qualify the tube expansion procedures prior to use in nuclear facilities. These tests include mockups for pneumatic and helium leak testing, and hydrostatic testing of mockups with rolled test joints at over 600 psi. Examples will include testing with 3 roll expanders compared to 5-roll expanders in heavy gauge titanium tubes, test results with a double tubesheet arrangement, and high strength joints with thin wall titanium tubes in brass tubesheets. An example comparing the pullout forces determined with a HEI beam strip analysis to the actual pullout test data is also discussed.
Experimental Investigation on Heat Transfer for Two-Phase Flow Under Natural Convection
Authors: Amižić Milan , Guyez Estelle , Seiler Jean-Marie
Published: October 2013

In the frame of severe accident research for the second and the third generation of nuclear power plants, some aspects of the concrete cavity ablation during the molten corium–concrete interaction are still remaining issues. The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt-through.

For the purpose of experimental investigation of thermal-hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched jointly by CEA, EDF, IRSN, GDF-Suez and SARNET. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the smallest pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s. This paper presents some preliminary conclusions deduced from the experiments which involve a liquid pool with the gas injection only from the bottom plate. A comparison with existing models for the assessment of heat transfer has also been carried out.
Dissociation of trimethylgallium on the ZrB2(0001) surface
Authors: Manandhar Kedar , Trenary Michael , Otani Shigeki , Zapol Peter
Published: October 2013

X-ray photoelectron spectroscopy and reflection absorption infrared spectroscopy (RAIRS) have been used to study the dissociative adsorption of trimethylgallium (TMG) on the ZrB2(0001) surface. Spectra were obtained as a function of annealing temperature following TMG exposure at temperatures of 95 and 300 K, and also as a function of TMG exposure for a surface temperature of 300 K. After annealing above 220 K, a significant decrease in the relative concentration of carbon and gallium occurred accompanied by a shift of ∼0.2 eV in the Ga 2p3/2 binding energy. The RAIR spectra show that after annealing to ∼220 K, only one CH3 deformation band at 1196 cm−1 remains, the intensity of which is considerably decreased indicating loss of at least one methyl group from TMG. Further annealing leads to the sequential loss of the other methyl groups. The first methyl desorbs while the last two dissociate to deposit two C atoms per TMG molecule onto the ZrB2 surface.
Can We Measure Autism?
Authors: Kohane Isaac S. , Eran Alal
Published: October 2013

Newly released definitions of autism demonstrate the need for precise diagnoses informed by a patient-information commons.
Foiling Inflammatory Bowel Disease
Authors: Mahfuz Mustafa
Published: October 2013

Aluminum enhances intestinal inflammation in a mouse model.
Out with the New, In with the Old
Authors: Narla Goutham
Published: October 2013

Computational 21st-century techniques pinpoint tricyclic antidepressants—discovered in the 1950s—as possible cancer therapeutics.
Pitfalls in Identifying Causal Mutations for Disease
Authors: Neale Benjamin M.
Published: October 2013

Current methods for determining deleterious mutations, even for known variants in well-established genes, need validation to confirm the phenotypic effects.
The ABC-Type Efflux Pump MacAB Protects Salmonella enterica serovar Typhimurium from Oxidative Stress
Authors: Bogomolnaya Lydia M. , Andrews Katharine D. , Talamantes Marissa , Maple Aimee , Ragoza Yury , Vazquez-Torres Andres , Andrews-Polymenis Helene
Published: October 2013

Multidrug efflux pumps are integral membrane proteins known to actively excrete antibiotics. The macrolide-specific pump MacAB, the only ABC-type drug efflux pump in
Salmonella
, has previously been linked to virulence in mice. The molecular mechanism of this link between
macAB
and infection is unclear. We demonstrate that
macAB
plays a role in the detoxification of reactive oxygen species (ROS), compounds that salmonellae are exposed to at various stages of infection.
macAB
is induced upon exposure to H
2
O
2
and is critical for survival of
Salmonella enterica
serovar Typhimurium in the presence of peroxide. Furthermore, we determined that
macAB
is required for intracellular replication inside J774.A1 murine macrophages but is not required for survival in ROS-deficient J774.D9 macrophages.
macAB
mutants also had reduced survival in the intestine in the mouse colitis model, a model characterized by a strong neutrophilic intestinal infiltrate where bacteria may experience the cytotoxic actions of ROS. Using an Amplex red-coupled assay,
macAB
mutants appear to be unable to induce protection against exogenous H
2
O
2
in vitro
, in contrast to the isogenic wild type. In mixed cultures, the presence of the wild-type organism, or media preconditioned by the growth of the wild-type organism, was sufficient to rescue the
macAB
mutant from peroxide-mediated killing. Our data indicate that the MacAB drug efflux pump has functions beyond resistance to antibiotics and plays a role in the protection of
Salmonella
against oxidative stress. Intriguingly, our data also suggest the presence of a soluble anti-H
2
O
2
compound secreted by
Salmonella
cells through a MacAB-dependent mechanism.

IMPORTANCE
The ABC-type multidrug efflux pump MacAB is known to be required for
Salmonella enterica
serovar Typhimurium virulence after oral infection in mice, yet the function of this pump during infection is unknown. We show that this pump is necessary for colonization of niches in infected mice where salmonellae encounter oxidative stress during infection. MacAB is required for growth in cultured macrophages that produce reactive oxygen species (ROS) but is not needed in macrophages that do not generate ROS. In addition, we show that MacAB is required to resist peroxide-mediated killing
in vitro
and for the inactivation of peroxide in the media. Finally, wild-type organisms, or supernatant from wild-type organisms grown in the presence of peroxide, rescue the growth defect of
macAB
mutants in H
2
O
2
. MacAB appears to participate in the excretion of a compound that induces protection against ROS-mediated killing, revealing a new role for this multidrug efflux pump.
A Critical Role for the Putative NCS2 Nucleobase Permease YjcD in the Sensitivity of Escherichia coli to Cytotoxic and Mutagenic Purine Analogs
Authors: Kozmin Stanislav G. , Stepchenkova Elena I. , Chow Stephen C. , Schaaper Roel M.
Published: October 2013

The base analogs 6-
N
-hydroxylaminopurine (HAP) and 2-amino-HAP (AHAP) are potent mutagens in bacteria and eukaryotic organisms. Previously, we demonstrated that a defect in the
Escherichia coli ycbX
gene, encoding a molybdenum cofactor-dependent oxidoreductase, dramatically enhances sensitivity to the toxic and mutagenic action of these agents. In the present study, we describe the discovery and properties of a novel suppressor locus,
yjcD
, that strongly reduces the HAP sensitivity of the
ycbX
strain. Suppressor effects are also observed for other purine analogs, like AHAP, 6-mercaptopurine, 6-thioguanine, and 2-aminopurine. In contrast, the
yjcD
defect did not affect the sensitivity to the pyrimidine analog 5-fluorouracil. Homology searches have predicted that
yjcD
encodes a putative permease of the NCS2 family of nucleobase transporters. We further investigated the effects of inactivation of all other members of the NCS2 family, XanQ, XanP, PurP, UacT, UraA, RutG, YgfQ, YicO, and YbbY, and of the NCS1 family nucleobase permeases CodB and YbbW. None of these other defects significantly affected sensitivity to either HAP or AHAP. The combined data strongly suggest that YjcD is the primary importer for modified purine bases. We also present data showing that this protein may, in fact, also be a principal permease involved in transport of the normal purines guanine, hypoxanthine, and/or xanthine.

IMPORTANCE
Nucleotide metabolism is a critical aspect of the overall metabolism of the cell, as it is central to the core processes of RNA and DNA synthesis. At the same time, nucleotide metabolism can be subverted by analogs of the normal DNA or RNA bases, leading to highly toxic and mutagenic effects. Thus, understanding how cells process both normal and modified bases is of fundamental importance. This work describes a novel suppressor of the toxicity of certain modified purine bases in the bacterium
Escherichia coli
. This suppressor encodes a putative high-affinity nucleobase transporter that mediates the import of the modified purine bases. It is also a likely candidate for the long-sought high-affinity importer for the normal purines, like guanine and hypoxanthine.