bluuebunny/crossref_metadata_2025 huggingface dataset:
Investigation of Quench-16 Experiment With MELCOR Computer Code
Authors: Vryashkova Petya , Groudev Pavlin , Stefanova Antoaneta
Published: November 2014

This paper presents a comparison of MELCOR calculated results with experimental data for the QUENCH-16 experiment. The analysis for the air ingress experiment QUENCH-16 has been performed by INRNE. The calculations have been performed with MELCOR code. The QUENCH-16 experiment has been performed on 27-th of July 2011 in the frame of the EC-supported LACOMECO program. The experiments have focused on air ingress investigation into an overheated core following earlier partial oxidation in steam. QUENCH-16 has been performed with limited pre-oxidation and low air flow rate. One of the main objectives of QUENCH-16 was to examine the interaction between nitrogen and oxidized cladding during a prolonged period of oxygen starvation. The bundle is made from 20 heated fuel rod simulators arranged in two concentric rings and one unheated central fuel rod simulator, each about 2.5 m long. The tungsten heaters were surrounded by annular ZrO2 pellets to simulate the UO2 fuel. The geometry and most other bundle components are prototypical for Western-type PWRs. To improve the obtained results it has been made a series of calculations to select an appropriate initial temperature of the oxidation of the fuel bundle and modified correlation oxidation of Zircaloy with MELCOR computer code. The compared results have shown good agreement of calculated hydrogen and oxygen starvation in comparison with test data.
Characterization of In-Core Water Chemistry for Corrosion Control of LWRs
Authors: Saji Genn
Published: November 2014

This paper updates scientific bases of water chemistry in applying the author’s recent theory, which integrates the elemental radiation- and electro-chemistry reactions in the “Butlar-Volmer equation,” presented in ICONE21-16525.

For the past several years the author has been trying to establish that the “long-cell” (a kin to macro-cell) corrosion mechanism is inducing practically all sorts of accelerated corrosion phenomena widely observed in water-cooled reactors, especially in aged plants.

The theoretical electrochemical potential differences have been benchmarked with the published in-pile test results for both PWR- and BWR water chemistry environments. However the author’s previous verification efforts were limited to the extent that the curves were fitted with experimental results at a single point. The author re-formulated the basic theory and found that the redox potential difference consists of an electrochemical part (e.g., Nernst equation of dissolved hydrogen or oxygen) and radiation-induced perturbation term, the latter diminishes to zero without radiation.

The author continued his studies to clarify whether our current scientific knowledge is sufficient to explain the in-core “chemistry” to reproduce the experimental results without the fitting parameter. Through his study he realized that the basic mechanism of the potential difference is still not sufficiently known.

No fitting parameter was used for the PWR water chemistry in the DH region for practical engineering applications, although it is indispensable to confirm the results with an in-pile test loop. In the BWR-NWC the theoretical redox potential out of core was still necessary to be fitted with the experimental results, due to an effect of residual hydrogen peroxide detected by the reference electrode. In addition the calculated potential shift is several times larger than the experimental observation. With the reformulation the scientific validity of the author’s theory is further confirmed. He believes that there is no doubt that the “long-cell” takes place in LWRs, although details are still debatable.
Molten Salt Fast Reactor Blanket Design and Proliferation Resistance Assessment
Authors: Bykov Valentyn , Křepel Jiři , Pautz Andreas
Published: November 2014

Molten Salt Reactor (MSR) designs are frequently accompanied by a blanket salt. This way the irradiation of the outer reactor wall will be strongly reduced. On the other hand, the barrier between the core and blanket will undergo higher irradiation and it will be necessary to replace it several times during the reactor lifetime. Furthermore, this blanket salt will also have a positive impact on neutron economy by improving the breeding performance.

In this paper a blanket of a generic two fluid molten salt reactor utilizing fast thorium-uranium cycle was investigated. This was done by tracking the evolution of uranium, neptunium and plutonium isotopes with burnup, which was then influenced by removal of uranium from the blanket. A significant reduction in the production of minor actinides was observed.

The uranium vector removed from the core was then investigated for proliferation resistance, using NUREC proliferation resistance metric and comparison with other weapon designs. The evaluation concluded that while the presence of U-232 increases radiological hazard associated with this uranium, thereby erecting a radiological barrier, it cannot be treated as “self-protecting” based on IAEA and NRC standards, requiring 1 Sv/h at 1m dose rate. Moreover ideas on how an interested party could reduce this radiological hazard were discussed.
Cross Sections Influence on Monte Carlo Based Burnup Codes
Authors: Chersola Davide , Lomonaco Guglielmo , Mazzini Guido
Published: November 2014

This paper reports the results of a comparison among JEFF and ENDF/B datasets when used by SERPENT and MONTEBURNS codes on a GFR-like configuration. Particularly, it shows a comparison between the two Monte Carlo based codes, each one adopting three different cross sections dataset, namely JEFF-3.1, JEFF-3.1.2 and ENDF/B-VII.1. Calculations have been carried out on the Allegro reactor, i.e. an experimental GFR-like facility that should be built in EU as GFR demonstrator. Results concern nuclear parameters as effective multiplication factor and fluxes, as well as the atomic densities for some important nuclides versus burnup.
Parametric Lattice Study of a Graphite-Moderated Molten Salt Reactor
Authors: Hombourger Boris A. , Křepel Jiři , Mikityuk Konstantin , Pautz Andreas
Published: November 2014

This article illustrates the influence of heterogeneity in an infinite lattice of a Molten Salt Reactor moderated by graphite. For a complete description of heterogeneity in a 2D lattice, two variables are needed; in this study the salt share in the unit cell and the channel radius are used. The equilibrium Thorium-based closed-cycle fuel composition is systematically derived for each chosen combination of points, and results such as kinf and the actinide vector composition are calculated. Results show that the heterogeneity effect can indeed be important for optimization of the core design of moderated molten salt reactors.
A Study on Weld Residual Stress Relaxation by Local Post Weld Heat Treatment for Circumferential Weld
Authors: Lee Seung-gun , Son Youngho
Published: November 2014

Weld residual stress is a troublesome problem in nuclear power plant, because it can accelerate crack growth in weld region. For low alloy steel, Post Weld Heat Treatment (PWHT) is essentially needed to relieve residual stress and to temper the hard regions in the heat affected zone (HAZ). Local PWHT is used when it is impractical to heat the whole component in a furnace. The rules and practices of related codes and standards, such as ASME and AWS, associated with local PWHT are quite different. For example, according to ASME Section III, the minimum width of heated band at each side of the weld shall be the thickness of the weld or 2 in., whichever is less. While, according to ASME B31.1, the width of heated band shall be at least three times the wall thickness at the weld of the thickest part being joined.

In this paper, the status of the related code and standard associated with local PWHT is briefly summarized, and baseline information on local PWHT is explained based on FEA (Finite Element Analysis) results and optimized local PWHT parameter is suggested to support current code of practices.
Prediction of Pressure Propagation in Piping Systems After ISLOCA
Authors: Yamada Takeshi , Hisamochi Kohei , Takii Taichi , Yasuda Kenichi
Published: November 2014

Interface-System Loss-of-Coolant Accident (ISLOCA) occurs by failure of isolation valves in piping systems. In this study, transient behavior of pressure propagation in piping systems after ISLOCA has been estimated. At first, capability of TRACG code to predict pressure propagation in a simple pipe line, which has expansion, contraction and bifurcation, has been investigated. Then, sensitivity analysis of valve opening period has been performed to investigate the behavior of pressure propagation in a simple pipe line. Finally, pressure propagation inside piping systems after ISLOCA has been simulated for Hitachi-GE standard Advanced boiling water reactor (ABWR) plant. Maximum pressure inside High Pressure Core Flooder (HPCF) systems has been less than 7.8MPa. TRACG code has been shown to be useful to predict transient behavior of pressure propagation in complicated network of piping systems after ISLOCA.
Simulation of Natural Convection Experiment Performed in MISTRA Containment Experimental Facility
Authors: Debevc Martin , Kljenak Ivo
Published: November 2014

An experiment on atmosphere mixing in a nuclear power plant containment experimental facility, which involved natural convection, was simulated with a CFD computer code. The experiment was performed in the MISTRA cylindrical vessel at CEA in Saclay (France). It involved gradual heating of structures near the vertical walls to induce buoyant flow and cause mixing of the atmosphere and breaking up of a previously established helium layer in the vessel upper region. The experiment was simulated with the CFD code ANSYS CFX. The simulation results are presented and discussed.
MTC 3D: A Fully Coupled Methodology for MSLB Safety Analysis
Authors: Royère Christian , Gonnet Michel , Manchid Brahim
Published: November 2014

The main steam line break (MSLB) is an overcooling accident that may lead to an over criticality, and so to a power increase, after the reactor trip. The most penalizing single failure is a RCCA bank stuck out of the core when the reactor trip occurs. This configuration leads to a strong asymmetry of the radial power shape combined with a strong asymmetry of the core inlet temperature that results in a strongly distorted 3D power distribution.

In the original design, the MSLB accident was studied with a simplified and conservative 0D method. The point kinetics approach requires the use of extremely conservative assumptions in order to account for the asymmetry in the core region that takes place during the transient. The use of the coupling between a three-dimensional neutronic code (SMART), a 3D core thermal-hydraulic code (FLICA cf. ref [4]) and a reactor coolant system code (MANTA cf. ref [3]) allows representing the 3D heterogeneity of the power shape and also of the resulting cross flows. In addition, this coupling allows determining moderator and Doppler feedback effects in a much more realistic way thus limiting accident consequences estimated.

A methodology, called MTC3D (for Méthode Totalement Couplée 3D in French), has been developed using the coupling between the three codes to perform the MSLB analysis. The physical dominant parameters of the transient are identified through a comprehensive sensitivity analysis. Then, a deterministic approach is used in the entire transient simulation considering dominant parameters in a penalizing way.

In a first step, neutronic data are determined with SMART calculations. In a second step, MANTA/SMART/FLICA transients are performed with penalized neutronic and thermal-hydraulic data. In a third step, as the steam line break transient is a relatively slow transient, the core power distribution is evaluated with a steady state SMART/FLICA calculation without penalization. In a last step, safety criteria, such as minimum DNBR (Departure from Nucleate Boiling Ratio) are calculated with FLICA calculations based on core power distribution calculated at the third step and boundary conditions calculated at the second step.

The use of 3D neutronic and detailed thermal-hydraulic codes to model the reactor core allows considering a more physical representation of the core configuration for transient analysis. The coupling between 3D neutronic and core thermal-hydraulic codes allows exhibiting intrinsic margins without over penalizations related to a simplified 0D method.
Low-Level Health Effects of Radiation: Nuclear Engineering Obligations
Authors: Kunze Jay F. , Sandquist Gary M.
Published: November 2014

It has been more than 30 years since the publication of data on health effects of low level began to appear in the scientific literature. Now, this extensive data bank clearly shows that the long-taught and utilized “Linear No-Threshold” (LNT) hypothesis is invalid and misleading at levels below 0.2 Sv (20 REM) per year. Below these levels, health and longevity are actually improved. Yet the imbedded and pervasive fear of low levels of radiation has resulted in planned and implemented nuclear plant shutdowns and eventual decommissioning, and has even affected the medical community regarding inappropriate concerns about dangers to patients receiving CT diagnostic imaging. These trends are now so serious and consequential to the future of nuclear energy and nuclear applications, that it is time for the nuclear engineering community to take an active role to dismiss the LNT hypothesis as incorrect and completely misleading with regard to low levels of radiation exposures to the public and to nuclear personnel. The ALARA concept and regulatory burden needs to be abandoned, and the realistic nuclear personnel exposure limits of the 1950 era should be re-adopted, and even increased.
Advanced Alternatives of VVER-440 Fuel Assembly
Authors: Tomaškovič Juraj , Dařílek Petr , Zajac Radoslav , Nečas Vladimír
Published: November 2014

The main goals of fuel development for pressurized water reactor are effectiveness and economic efficiency. Both requirements can be achieved by gradual increase of discharged fuel burn-up and prolongation of fuel cycle. The mentioned effects can be reached by optimisation of fuel assembly profiling, fuel enrichment raise, and by parasitic absorption reduction. These methods were used in VVER-440 fuel assembly optimisation, described in this paper. Fuel pin configurations with enrichment limit 5 % and also enlarged one up to 5.95 % U235 were designed. Reduction of parasitic absorption was limited by carcass frame of the assembly. Basic characteristics of the best assembly proposals are presented and effects on equilibrium fuel cycle of VVER-440 reactor are characterized.
Ageing Management of Dry Storage Systems for Long-Term Storage
Authors: Garcia Justo , Issard Hervé
Published: November 2014

Nuclear reactors generate used nuclear fuel, which needs to be safely managed. After discharge from operating nuclear reactors, the used nuclear fuel stored during a certain period of time in reactor pool is removed for reprocessing, disposal, or storage elsewhere.

A need for additional storage exists because of delays in used nuclear fuel disposition decisions and the continually increasing volume of used nuclear fuel discharged from reactors. Interim storage of used nuclear fuel is an essential part of the Nuclear Fuel Cycle providing nuclear safety and ensuring protection to the public.

Interim storage of used nuclear fuel currently used includes:

• Dry storage in vaults, casks and containers

• Wet storage, in silos, pools, outside reactor operating areas

Interim dry storage systems were originally designed for limited periods, generally 40–60 years. Interim dry storage is safe as shown by important industrial feedback and the operational records, even from severe accidents.

With extended interim dry storage, technical and safety investigations are presently being carried out concerning used nuclear fuel behavior, and the storage structures and their system components to demonstrate and justify the ability of the systems to store safely and securely the used nuclear fuel for a much longer period of time.

These investigations are part of the ageing management program for the storage system which should address a description of the ageing that could adversely affect structures and components important to safety.

This paper provides highlights of issues related to managing effects on dry cask storage systems for long-term interim storage and transportation of used nuclear fuel.
Development of Inherently Safe Technologies for Large Scale BWRs: (4) Hydrogen Explosion Prevention System Using SiC Fuel Claddings
Authors: Ishibashi Ryo , Ikegawa Tomohiko , Noshita Kenji , Kitou Kazuaki , Kamoshida Mamoru
Published: November 2014

In the aftermath of the lessons learned from the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), including a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents.

One of inherently safe technologies currently under development is a system to prevent hydrogen explosion during severe accidents (SAs). This hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding of silicon carbide (SiC), and a passive autocatalytic recombiner (PAR). Replacing the zircaloy (Zry) claddings currently used in LWRs with the SiC claddings decreases the hydrogen generation and thus decreases the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B) such as an operation floor. The PAR recombines the leaked hydrogen gas so as to maintain the hydrogen concentration at less than the explosion limit of 4 % in the R/B.

The advantages of using SiC claddings in the system were examined through experiments and SA analysis. Results of steam oxidation tests confirmed that SiC was estimated to show 2 to 3 orders of magnitude lower hydrogen generation rates during oxidation in a high temperature steam environment than Zry. Results of SA analysis showed that the total amount of hydrogen generation from fuels was reduced to one fifth or less. Calculation also showed that the lower heat of the oxidation reaction of SiC moderated the steep generation with the temperature increase. We expected this moderated steep generation to reduce the pressure increase in the PCV as well as prevent excess amounts of leaked hydrogen from hydrogen disposal rate capacity using PARs.

The SiC cladding under consideration consists of an inner metallic layer, a SiC/SiC composite substrate, and an outer environment barrier coating (EBC). A thin inner metallic layer in combination with a SiC/SiC composite substrate functions as a barrier for fission products. EBC is introduced to have both corrosion resistance in high temperature water environments during normal operation and oxidation resistance in high temperature steam environments during SA. Further reduction of the hydrogen generation rate in high temperature steam by improving the EBC is expected to decrease the total amount of hydrogen generation even more.
Technical Basis of Fatigue Crack Growth Rate Curve for Ni-Base Alloy in BWR Water Environment
Authors: Ogawa Takuya , Itatani Masao , Saito Toshiyuki , Nagase Hiroshi , Aoike Satoru , Yoneda Hideki
Published: November 2014

When the flaws are detected in Japanese nuclear power components by in-service inspection, structural integrity assessment are performed in the technical judgment on continuous service. If cyclic loading is assumed, fatigue crack growth analysis should be conducted based on the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code). However, fatigue crack growth analysis for BWR components consisting of Ni-base alloy is currently impossible, since the reference curve of fatigue crack growth rate for Ni-base alloy in BWR water environment is not yet prescribed in the JSME FFS Code.

In this study, fatigue crack growth behavior of Ni-base alloy used for Japanese BWR plants in BWR water environment was investigated. Based on the experimental data, the fatigue crack growth rate curve was evaluated. Four test parameters of material, corrosion potential, stress ratio and load rising time were considered. As a result of fatigue crack growth tests, the effects of all test parameters on the fatigue crack growth behavior were found. A Mean curve of fatigue crack growth rate in Paris law format, which was a function of stress ratio and rising time, was formulated based on crack growth data in normal water chemistry (corrosion potential was over 150 mVSHE) for weld metal and heat affected zone (HAZ), respectively. A reference curve of fatigue crack growth rate was also formulated by the statistical treatment considering the scatter of crack growth rate.

Further, in order to determine the threshold stress intensity factor range ΔKth of reference curve of fatigue crack growth, ΔK decreasing tests were conducted under the test condition of 1 second of rising time. As a result, the threshold value of ΔK was evaluated based on the ASTM E 647, and the ΔKth of the reference curve was conservatively determined considering the margin.
Accident at Fukushima Dai-Ichi Nuclear Power Plant: Lessons Learned for the Czech Republic
Authors: Vesely Ladislav , Dostal Vaclav
Published: November 2014

Accident at Fukushima Dai-Ichi nuclear power plant significantly affected the nuclear industry at time when everybody was expecting the so called nuclear renaissance. There is no question that the accident has at least slowed it down. Research into this accident is taking place all over the world. In this paper we present the findings of research on Fukushima nuclear power plant accident in relation to the Czech Republic. The paper focuses on the analysis of human performance during the accident. Lessons learned from the accident and main human errors are presented. First the brief factors affecting the human performance are discussed. They are followed by the short description of activities on units 1–3. The key human errors in the accident mitigation are then identified. On unit 1 the main error is wrong understanding and operation of isolation condenser. On unit 2 the main errors were unsuccessful depressurization with subsequent delay of coolant injection. On unit 3 the main error is the shutdown of high pressure cooling injection system without first confirming that different means of cooling are available. These errors lead to fuel damage. On unit 1 the fuel damage was probably impossible to prevent, however on unit 2 and 3 it could be probably prevented. The lessons learned for the Czech Republic were presented. They can be summarizes as follows: be sure that plant personnel can and knows how to monitor and operate the crucial plant components, be sure that the procedures on how to fulfill the critical safety functions are available in the symptomatic manner for situations when there is no power available at the plant, train personnel for these situations and have sufficient human resource available for these situations.
Water Chemistry Sensitivity on Fibrous Debris Bypass Through a Containment Sump Strainer
Authors: Lee Saya , Hassan Yassin A. , Vaghetto Rodolfo , Abdulsattar Suhaeb , Kappes Matthew
Published: November 2014

Fibrous debris generated from insulation materials during a Loss of Coolant Accident (LOCA) might be transported to the containment sump strainer. Once the fibrous debris reaches the sump strainer, it might accumulate on the strainer causing loss of Net Positive Suction Head (NPSH), called the upstream effect, or it might penetrate through the strainer becoming a source of clogging of flow channel in the core (in-vessel effect). This is the main idea of the Generic Safety Issue (GSI) 191 initiated by the U.S. Nuclear Regulatory Commission (U.S.NRC). In this study, the authors focused on the quantity of the fibrous debris bypass through a containment sump strainer which eventually might cause the in-vessel effect. In order to produce the fibrous debris, a method suggested by the Nuclear Energy Institute (NEI) was applied to one-side-baked NUKON fiberglass insulation materials. The quantity and characteristics of debris penetrating through the strainer may be affected by the properties of water such as pH and electrical conductivity (EC). Total quantity of the debris bypass for a certain period was measured and the effects of the water chemistry were observed. At higher pH and lower EC more fibrous debris penetrated the strainer.
Evaluation of Safety Performances of a Superficial Disposal Facility Subject to an Aircraft Impact and Fuel Burning
Authors: Stefanini Lorenzo , Lo Frano Rosa , Bertocchi Fulvio , Forasassi Giuseppe
Published: November 2014

Safety and security aspects are of meaningful importance in the design of nuclear facilities.

In this study, the attention is so focused on the potential damaging effects that a large civilian airplane impact could bring in safety relevant structures, like a superficial repository similar to El Cabril one.

Safety performances of such a type of superficial disposal facility, subjected to the aircraft impact and fuel burning, have been analysed and discussed.

Conservative assumptions have been made: normal impact on the lateral repository surface, fire scenario based on the amount of fuel burnt.

Load functions (calculated with the Riera approach) and the maximum temperature reached by fuel during its combustion were used as input (boundary condition) in the numerical simulations as well as the damaging phenomena occurring in the concrete structure.

Numerical analyses, by MARC© code, allowed to simulate the thermo-structural performances of the superficial repository.

The obtained results showed that a repository wall thickness, ranging from 0.6 to 0.9 m, is not sufficient to prevent the penetration of wall itself. Despite the ongoing concrete degradation phenomena, the global strength of the repository seemed to be guaranteed.
Horizontal Drop Test Evaluation of a Packaging System
Authors: Sanfiorenzo Antonio , Lo Frano Rosa , Pugliese Giovanni , Forasassi Giuseppe
Published: November 2014

In the transportation of radioactive waste, the package is designed as the major engineered system capable to ensure the containment and provide safety functions, such as radiation shielding, structural integrity against external mechanical and thermal loads, dissipation of the decay heat, etc. Packaging systems are designed in accordance to rigorous acceptance requirements, like the International Atomic Energy Agency (IAEA) ones, so to provide protection to human being and environment against radiation exposure and contamination, particularly in reference accident scenarios including, as it is widely known in literature, drop, puncture, fire and submersion tests. The scope of the present study is to evaluate the structural response and performance in a free drop test condition of a new Italian packaging system that should be used for the transportation of low and intermediate level radioactive wastes. For this purpose the carried out numerical analyses are presented and discussed. The numerical analyses, performed by the finite element MARC® code, simulate the behaviour of the packaging system components: the overpack, gasket, cover lid, bolts and a concrete matrix representative of the radioactive content.

The obtained results for 1.2 m horizontal drop, on a flat and unyielding surface, were critically analysed and also compared to the experimental ones obtained from the experimental test campaign performed at the Unipi test facility on the new Italian packaging system considered.

The stress and acceleration values indicate that the package, although rather local deformations in correspondence of bolts and secondary lid, is capable to withstand the dynamic loading generated during the drop test without any unacceptable loss of the safety features.
Data Reconciliation of the Turbine Section: Evaluation of Estimation Uncertainty
Authors: Saarela Olli , Wingstedt Emil
Published: November 2014

Data reconciliation is technique for reducing measurement uncertainty by adjusting measured data to comply with a first-principles process model, most importantly with mass and energy balances. It also provides estimates for modelled unmeasurable process variables and estimates for the uncertainties of the computed values. For computing these estimates the process model has to include estimates of measurement uncertainties defined a priori. A priori consideration of all potential sources of uncertainty is far from trivial. This paper discusses a data-driven approach of uncertainty evaluation, based on identifying and subtracting variability modes affecting multiple measurements. Possible bias in the measurements is not considered. The approach is applied to evaluate the uncertainties of estimates computed with a data reconciliation model of a turbine section of a nuclear power plant.
Conjugated Numerical Study on the Performance of Microchannel Heat Sink Using Al2O3/H2O Nanofluid
Authors: Wu J. M. , Zhao J. Y.
Published: November 2014

High power electronics are widely used in many different areas such as integrated circuit (IC) boards in nuclear reactor control system. Thermal management of electronic devices has been a topic of great interest among many researchers over the last few decades. Microchannel is one of several high-heat-flux removal techniques. Nanofluids with enhanced thermal conductivity and strong temperature- and size-dependent thermal properties are expected to be utilized in microchannels as coolants, which leads to a promising future for such high-heat-flux systems as cooling systems. The performance of the microchannel heat sink (MCHS) using water and Al2O3/water nanofluids, with consideration of different substrate materials, is numerically investigated and compared in the present paper to identify the combined effects of working fluids and substrate materials on the thermal resistance, pumping power and temperature distribution on the substrate surface of a heat sink.
Replacement of I&C System for the Four Main Pumps in the Fuel Channels Cooling Circuit of “Maria” Polish Research Reactor
Authors: Drobnik Rafał , Głowacki Andrzej
Published: November 2014

In this work the replacement of the I&C connected with the four main pumps in a Polish research reactor called “Maria” is discussed. In particular, this paper focuses on the aspect of the whole system of pump parameters measurements (temperature, vibrations, power) and pump control system. The whole architecture of the I&C system (including power supply, redundancy of electrical and control system) will be showed. During the I&C replacement some problems arose, such as the change in the pump control system from analog to digital system; the link existing in the reactor part of the measurement made in analogue to a digital system, as part of the measurements, which were carried out in an old control system; the code for the PLC controller program dedicated exclusively to the requirements of control systems pumps. All of these problems will be discussed here, together with the proposed solutions. Moreover, I&C functional and operational tests carried out after the placement of the pumps will be described, such as loss of power and failure of one PLC controller.
Developing Structural Design Criteria for Fusion Reactor In-Vessel Components
Authors: Kalsey M. , Porton M.
Published: November 2014

Co-ordinated international activities are underway to complete an EU fusion demonstration power plant conceptual design by 2020, for construction in the early 2030s. However, many engineering design challenges still remain. For example, the design of in vessel components is technically challenging, with many design constraints needing to be mutually satisfied. One of the key challenges is the design validation of in-vessel structural components operating within the unique fusion environmental conditions. It has been recognized that these components would benefit from design criteria that have been specifically developed, providing improved accuracy and appropriate conservatism. This would provide DEMO designers with a much needed increase in design space and guidance.

This paper highlights the justification behind developing fusion specific structural design criteria, discusses the key technical gaps of existing criteria and elaborates on the need to develop non-technical aspects of the documentation including consideration of human factors in design validation. Finally, a structured design criteria development plan is presented, highlighting planning work by the European fusion community to address these gaps in step with the current design milestones for the realisation of an EU DEMO.
Simulation of PWR Crud
Authors: Byers W. A. , Wang G. , Young M. Y. , Deshon J.
Published: November 2014

Corrosion products deposit on Pressurized Water Reactor (PWR) fuel rod cladding surfaces and can create a number of issues including increased cladding temperature, elevated cladding corrosion, and the precipitation of boron species within the deposits. The deposits can also release and lead to increased radiation fields on system ex-core surfaces. These effects can vary widely from plant-to-plant. The amount of the deposits, commonly known as crud, is an important factor in determining the impact, but other parameters such as crud thickness, porosity, and composition are also thought to be important.

The Electric Power Research Institute (EPRI) has sponsored a number of programs to better understand the characteristics of crud and its effects. Crud has been sampled by fuel scraping and by collecting suspended crud during operation and during fuel cleaning. The chemistry and structure of the crud was then characterized. These data were then used to create simulated crud in laboratory heated rod tests. These tests explored how the crud deposits affected heat transfer at the rod surface and the interaction between the crud and the simulated coolant.

This paper discusses the nature of PWR crud and some of the practical aspects of crud simulation. Different approaches to laboratory crud creation will be reviewed, and the success in matching plant crud characteristics will be shown, with special emphasis on the production of crud for thermal conductivity measurement.
{UO2}2+ Extraction Using Ionic Liquids in Intensified Extractors
Authors: Tsaoulidis Dimitrios , Plechkova Natalia , Seddon Kenneth R. , Angeli Panagiota
Published: November 2014

The extraction of dioxouranium(VI) from aqueous nitric acid solutions (3M) into TBP/ionic liquid mixtures (30% v/v), relevant to spent fuel reprocessing, was investigated experimentally in intensified small channel extraction units. The intensified extractor consisted of a “T” shaped inlet and a FEP channel of 0.5 mm internal diameter. Important hydrodynamic characteristics such as, plug length, and plug speed were determined by means of bright field imaging, and were related to the extraction performance of the extraction unit. The amount of dioxouranium(VI) extracted was measured online by using UV-Vis spectroscopy.
The Modern Technology of Radiation Oncology, Volume 3: A Compendium for Medical Physicists and Radiation Oncologists
Authors: Xia Ping
Published: November 2014

The Modern Technology of Radiation Oncology, Volume 3: A Compendium for Medical Physicists and Radiation Oncologists. Editor: J. Van Dyk, Medical Physics Publishing, Madison, WI, 2013. 604 pp. Price: $150.00. ISBN: 978‐1‐9305‐2457‐6 (hardcover).
Autotransporter-Based Antigen Display in Bacterial Ghosts
Authors: Hjelm Anna , Söderström Bill , Vikström David , Jong Wouter S. P. , Luirink Joen , de Gier Jan-Willem
Published: November 2014

Bacterial ghosts are empty cell envelopes of Gram-negative bacteria that can be used as vehicles for antigen delivery. Ghosts are generated by releasing the bacterial cytoplasmic contents through a channel in the cell envelope that is created by the controlled production of the bacteriophage ϕX174 lysis protein E. While ghosts possess all the immunostimulatory surface properties of the original host strain, they do not pose any of the infectious threats associated with live vaccines. Recently, we have engineered the
Escherichia coli
autotransporter hemoglobin protease (Hbp) into a platform for the efficient surface display of heterologous proteins in Gram-negative bacteria, HbpD. Using the
Mycobacterium tuberculosis
vaccine target ESAT6 (early secreted antigenic target of 6 kDa), we have explored the application of HbpD to decorate
E. coli
and
Salmonella
ghosts with antigens. The use of different promoter systems enabled the concerted production of HbpD-ESAT6 and lysis protein E. Ghost formation was monitored by determining lysis efficiency based on CFU, the localization of a set of cellular markers, fluorescence microscopy, flow cytometry, and electron microscopy. Hbp-mediated surface display of ESAT6 was monitored using a combination of a protease accessibility assay, fluorescence microscopy, flow cytometry and (immuno-)electron microscopy. Here, we show that the concerted production of HbpD and lysis protein E in
E. coli
and
Salmonella
can be used to produce ghosts that efficiently display antigens on their surface. This system holds promise for the development of safe and cost-effective vaccines with optimal intrinsic adjuvant activity and exposure of heterologous antigens to the immune system.
New Multilocus Variable-Number Tandem-Repeat Analysis Tool for Surveillance and Local Epidemiology of Bacterial Leaf Blight and Bacterial Leaf Streak of Rice Caused by Xanthomonas oryzae
Authors: Poulin L. , Grygiel P. , Magne M. , Gagnevin L. , Rodriguez-R L. M. , Forero Serna N. , Zhao S. , El Rafii M. , Dao S. , Tekete C. , Wonni I. , Koita O. , Pruvost O. , Verdier V. , Vernière C. , Koebnik R.
Published: November 2014

Multilocus variable-number tandem-repeat analysis (MLVA) is efficient for routine typing and for investigating the genetic structures of natural microbial populations. Two distinct pathovars of
Xanthomonas oryzae
can cause significant crop losses in tropical and temperate rice-growing countries. Bacterial leaf streak is caused by
X. oryzae
pv. oryzicola, and bacterial leaf blight is caused by
X. oryzae
pv. oryzae. For the latter, two genetic lineages have been described in the literature. We developed a universal MLVA typing tool both for the identification of the three
X. oryzae
genetic lineages and for epidemiological analyses. Sixteen candidate variable-number tandem-repeat (VNTR) loci were selected according to their presence and polymorphism in 10 draft or complete genome sequences of the three
X. oryzae
lineages and by VNTR sequencing of a subset of loci of interest in 20 strains per lineage. The MLVA-16 scheme was then applied to 338 strains of
X. oryzae
representing different pathovars and geographical locations. Linkage disequilibrium between MLVA loci was calculated by index association on different scales, and the 16 loci showed linear Mantel correlation with MLSA data on 56
X. oryzae
strains, suggesting that they provide a good phylogenetic signal. Furthermore, analyses of sets of strains for different lineages indicated the possibility of using the scheme for deeper epidemiological investigation on small spatial scales.
Dynamics of the Establishment of Multinucleate Compartments in Fusarium oxysporum
Authors: Shahi Shermineh , Beerens Bas , Manders Erik M. M. , Rep Martijn
Published: November 2014

Nuclear dynamics can vary widely between fungal species and between stages of development of fungal colonies. Here we compared nuclear dynamics and mitotic patterns between germlings and mature hyphae in
Fusarium oxysporum
. Using fluorescently labeled nuclei and live-cell imaging, we show that
F. oxysporum
is subject to a developmental transition from a uninucleate to a multinucleate state after completion of colony initiation. We observed a special type of hypha that exhibits a higher growth rate, possibly acting as a nutrient scout. The higher growth rate is associated with a higher nuclear count and mitotic waves involving 2 to 6 nuclei in the apical compartment. Further, we found that dormant nuclei of intercalary compartments can reenter the mitotic cycle, resulting in multinucleate compartments with up to 18 nuclei in a single compartment.
Spectrum of patients intentionally poisoned with an emulsified concentrate pendimethalin herbicide
Authors: Moon Jeongmi , Chun Byeongjo
Published: November 2014

Objective

Studies on the acute toxicity of pendimethalin herbicide in humans are limited. Therefore, this study investigated the clinical characteristics of acute intentional pendimethalin herbicide poisoning.

Material and method

A retrospective observational case series was conducted involving 17 patients with a history of intentional pendimethalin herbicide ingestion. Data were collected on clinical manifestations, management and final outcome.

Result

The mortality rate was 0%; however, four patients (23.5%) exhibited metabolic acidosis, hypotension or respiratory failure within the first 24 h after ingestion and required admission to the intensive care unit. The most common complication was respiratory failure, followed by hypotension. Complicated patients tended to show an altered mental state and X-ray abnormalities at presentation.

Discussion and conclusions

Physicians should be aware that patients who have been poisoned with pendimethalin herbicide, and particularly patients with a depressed mental state and X-ray abnormalities at presentation, may exhibit metabolic acidosis, hypotension, respiratory failure or pancreatitis.
KIT genetic alterations in anorectal melanomas
Authors: Santi Raffaella , Simi Lisa , Fucci Rossella , Paglierani Milena , Pepi Monica , Pinzani Pamela , Merelli Barbara , Santucci Marco , Botti Gerardo , Urso Carmelo , Massi Daniela
Published: November 2014

Background

Mucosal melanomas (MM) represent a heterogeneous tumour population that exhibits site-specific molecular profiles.

Aims

In a multicentre retrospective study, we investigatedKITaberrations in primary anorectal (AR) melanomas compared with melanoma metastatic to the gastrointestinal (GI) tract.

Methods

Primary AR MM (n=31) and GI metastatic melanoma (n=27) were studied forKITmutations on exons 11, 13, 17 and 18 by high-resolution melting analysis, direct sequencing and c-KIT expression by immunohistochemistry. Selected cases were also investigated for increasedKITgene copy number by fluorescent in situ hybridisation.

Results

FunctionalKITmutations were demonstrated in 11/31 (35.5%) of AR melanomas and in 1/26 (3.8%) of GI melanoma metastases (p=0.004). A significant difference emerged between primary and metastatic MM with regards to KIT-positive immunostaining (p=0.002). Immunohistochemical c-KITprotein overexpression did not correlate withKITmutational status. IncreasedKITcopy number was demonstrated in 5/20 AR primary cases.

Conclusions

The rate of functional mutations inKITis significantly higher in AR MM than in GI metastatic melanoma. KIT protein overexpression does not correlate withKITmutations and cannot be used for screening purposes. Recognising the molecular heterogeneity of MM helps to identify patients who require a different therapeutic approach.
Gas-breakthrough pressure of FEBEX bentonite
Authors: Gutiérrez-Rodrigo Vanesa , Villar María Victoria , Martín Pedro Luis , Romero Francisco Javier , Barcala José Miguel
Published: November 2014

The gas-breakthrough pressure values in saturated FEBEX bentonite were determined for different dry densities and sample sizes. They increased clearly with dry density and were always higher than the swelling pressure of the bentonite. In high-density samples, gas flow tended to stop abruptly after breakthrough, whereas, in lower density samples, gas flow decreased gradually until a given pressure gradient was reached. The permeabilities computed after breakthrough (which usually did not stabilize) were inversely related to dry density. This would indicate that, despite the fact that flow took place through preferential pathways, the bentonite matrix and its swelling conditioned the ease of pathway formation. These paths sometimes closed quickly after breakthrough and others remained open, allowing a gradual decrease in gas flow. After resaturation of the bentonite, the same breakthrough pressures and permeabilities were found, pointing to the perfect healing of these preferential pathways. A sealed interface along the bentonite did not seem to affect the breakthrough pressure or permeability values.
Transcriptional mechanisms and protein kinase signaling mediate organic dust induction of IL-8 expression in lung epithelial and THP-1 cells
Authors: Gottipati Koteswara R. , Bandari Shiva Kumar , Nonnenmann Matthew W. , Levin Jeffrey L. , Dooley Gregory P. , Reynolds Stephen J. , Boggaram Vijay
Published: November 2014

Exposure to the agricultural work environment is a risk factor for the development of respiratory symptoms and chronic lung diseases. Inflammation is an important contributor to the pathogenesis of tissue injury and disease. Cellular and molecular mechanisms mediating lung inflammatory responses to agricultural dust are not yet fully understood. We studied the effects of poultry dust extract on molecular regulation of interleukin-8 (IL-8), a proinflammatory cytokine, in A549 and Beas2B lung epithelial and THP-1 monocytic cells. Our findings indicate that poultry dust extract potently induces IL-8 levels by increasing IL-8 gene transcription without altering IL-8 mRNA stability. Increase in IL-8 promoter activity was due to enhanced binding of activator protein 1 and NF-κB. IL-8 induction was associated with protein kinase C (PKC) and mitogen-activated protein kinase (MAPK) activation and inhibited by PKC and MAPK inhibitors. IL-8 increase was not inhibited by polymyxin B or l-nitroarginine methyl ester, indicating lack of involvement of lipopolysaccharide and nitric oxide in the induction. Lung epithelial and THP-1 cells share common mechanisms for induction of IL-8 levels. Our findings identify key roles for transcriptional mechanisms and protein kinase signaling pathways for IL-8 induction and provide insights into the mechanisms regulating lung inflammatory responses to organic dust exposure.