bluuebunny/crossref_metadata_2025_split huggingface dataset:
Functional characterization of the chloroplast ferric chelate oxidoreductase enzyme
Authors: Solti Ádám , Müller Brigitta , Czech Viktória , Sárvári Éva , Fodor Ferenc 
Published: February 2014

Summary

Iron (Fe) has an essential role in the biosynthesis of chlorophylls and redox cofactors, and thus chloroplast iron uptake is a process of special importance. The chloroplast ferric chelate oxidoreductase (cFRO) has a crucial role in this process but it is poorly characterized.

To study the localization and mechanism of action of cFRO, sugar beet (Beta vulgaris cv Orbis) chloroplast envelope fractions were isolated by gradient ultracentrifugation, and their purity was tested by western blotting against different marker proteins. The ferric chelate reductase (FCR) activity of envelope fractions was studied in the presence of NAD(P)H (reductants) and FAD coenzymes. Reduction of Fe(III)‐ethylenediaminetetraacetic acid was monitored spectrophotometrically by the Fe(II)‐bathophenanthroline disulfonate complex formation.

FCR activity, that is production of free Fe(II) for Fe uptake, showed biphasic saturation kinetics, and was clearly associated only to chloroplast inner envelope (cIE) vesicles. The reaction rate was > 2.5 times higher with NADPH than with NADH, which indicates the natural coenzyme preference of cFRO activity and its dependence on photosynthesis.

FCR activity of cIE vesicles isolated from Fe‐deficient plants also showed clear biphasic kinetics, where the KM of the low affinity component was elevated, and thus this component was down‐regulated.
Anxiety and stress in mothers of food‐allergic children
Authors: Lau Gar‐Yen , Patel Nisha , Umasunthar Thisanayagam , Gore Claudia , Warner John O. , Hanna Heather , Phillips Katherine , Mohd Zaki Amirah , Hodes Matthew , Boyle Robert J. 
Published: February 2014

AbstractBackgroundPrevious reports suggest that parents especially mothers of food‐allergic children may have increased anxiety. Studies with an appropriate control group have not been undertaken, and the determinants of such anxiety are not known. We compared measures of anxiety and stress in mothers of food‐allergic children and atopic non‐food‐allergic children, with anxiety and stress in mothers of children with no chronic illness.MethodsCross‐sectional study of mothers attending a hospital appointment for their 8‐ to 16‐year‐old child. Mothers of children with food allergy, asthma but no food allergy or no chronic illness completed questionnaires including State‐Trait Anxiety Inventory, Perceived Stress Scale and measures of anxiety and psychologic adjustment in their child.ResultsForty mothers of food‐allergic children, 18 mothers of asthmatic children without food allergy and 38 mothers of children with no chronic illness (controls) were recruited. Mothers of food‐allergic children showed increased state anxiety – median anxiety score 38.0 (IQR 30.0, 44.0) food allergy, 27.0 (22.0, 40.0) control p = 0.012; and increased stress – median stress score 18.5 (12.0, 22.0) food allergy, 14.0 (7.5, 19.5) control p = 0.035. No significant differences were seen between mothers in the asthmatic group and controls. In multivariate analysis, previous food anaphylaxis (p = 0.008) and poorly controlled asthma (p = 0.004) were associated with increased maternal anxiety. Child anxiety and adjustment did not differ between food‐allergic and control groups.ConclusionsMothers of food‐allergic children have increased anxiety and stress compared with mothers of children with no chronic illness. Anaphylaxis and poorly controlled asthma are associated with maternal anxiety.
The Discussion of Nuclear Power Plant’s Cooling Chain Design for Freezing Site
Authors: Hu Jian , Jiang Xulun , Yang Ting , Xiao Wei , Wang Zhengguang 
Published: February 2014

In order to satisfy the IAEA safety requirement, the conservative design of Nuclear Power Plant’s (NPP) cooling chain is applied by consideration of maximum heat load and high ultimate heat sink temperature. However, for the freezing site, combined with the low heat load in cooling chain, the low service water temperature in winter may possibly causes the unavailability of cooling chain.
This paper is mainly focused on the once-through cooling chain design at freezing site. By analyzing the thoughts and applicability of different kinds of cooling chain design for anti-freezing consideration, one solution called “SEC thermal discharge reflux” is proposed. The solution has already been approved by National Nuclear Safety Administration (NNSA) in China, and applied in one of the northern China NPP.
Comparing with the traditional anti-freezing designs, the solution has its feature as “Reactor low power operation”, “Reactor first time start-up” and “Change-over of RRI/SEC trains in winter”. Besides, it is also able to prevent the heat exchanger of RRI side from icing and SEC side from sediment deposition and mussel growth. The solution is applicable for once-through cooling type NPP, and is not restricted by the reactor type.
Live Experimental Results of Melt Pool Behaviour in the PWR Lower Head With Insulated Upper Lid and External Cooling
Authors: Gaus-Liu Xiaoyang , Miassoedov Alexei 
Published: February 2014

In-vessel melt retention has drawn renewed concern as an important severe accident management measure in existing and advanced light water reactors. Despite numerous studies the central question whether the maximum heat flux in a melt pool could exceed the critical heat flux (CHF) is not fully answered. The uncertainty comes from the variety of accident scenarios and the corresponding melt pool configurations, as well as from the applicability of the experimental results to the reactor case. It is therefore necessary to examine the melt heat transfer under different pool configurations and cooling conditions, as well as to compare the experimental results coming from different test vessel geometries and cooling regimes.
This study investigates the heat transfer characteristics of an oxidic pool in the PWR lower plenum in the case when the vessel wall is externally cooled by water, and the melt upper surface is free in a closed insulated environment. Thus the melt pool cooling conditions are quasi-isothermal at the inclined sidewall and at the upper surface free surface with thermal radiation. This pool configuration can occur before the melt layer stratification begins or the melt pool is composed only of oxide melt under certain melt relocation sequences.
A non-eutectic nitrate mixture with the composition of 20% NaNO3−80% KNO3 in mole relation is used as the simulant melt. Besides the determination of melt temperature and heat flux in their global average values, emphasis are given on the characterization of the axial distribution of melt temperature and heat flux at different power densities and pool heights. Results obtained in hemispherical geometry are analyzed and compared with other studies conducted under similar boundary conditions. The characterization of the heat flux distribution provide important data for the prediction of the maximum heat flux in the reactor case with similar boundary conditions and the evaluation of the concept of in-vessel melt retention by external cooling.
Hybrid Feedforward and Feedback Control of Steam Temperature in a Once-Through Direct Cycle Canadian SCWR
Authors: Sun Peiwei , Zhang Jianmin 
Published: February 2014

Supercritical Water-cooled Reactor (SCWR) is the only water cooled reactor proposed for Generation IV potential nuclear reactors. Canadian SCWR is a pressure-tube type SCWR under development in Canada. A linearized dynamic model of the Canadian SCWR is developed based on the lumped-parameter method to investigate its dynamic characteristics. Through dynamic analysis, the Canadian SCWR can be modelled as a three-input and three-output system with high power-to-flow ratio and strong cross-coupling. Among the three outputs, the dynamics of the steam temperature are the most complex and vulnerable to disturbances, especially those from the reactor power. It is difficult and challenging to control the steam temperature.
In this paper, a hybrid feedforward and feedback control strategy is adopted for control system design of the steam temperature. Feedforward control can reduce the effect of the measured disturbance on the steam temperature. It starts to act before the disturbance actually affects the steam temperature. Feedback control is used to track the setpoint change and suppress unmeasured disturbances. The uncompensated effects from the feedforward control can also be eliminated by the feedback control. Combined feedforward and feedback control scheme can significantly improve performance.
The feedforward controller and feedback controller are designed based on the linearized dynamic model. The control performance is verified using nonlinear dynamic model. Through simulation, the steam temperature variation using designed control system satisfies its limit to a typical load pattern.
Study on the Prestressing Effect in Reinforced Concrete Containment by Using ANSYS
Authors: Sun Feng , Pan Rong 
Published: February 2014

Prestressed reinforced concrete containment is an important part among the safety-related buildings, and the presressing technology is widely used in containment construction which is best suited to the cylindrical-type containment topped by a dome. It has been recognized that traditional methods for analysis of containment prestressing system have been unable to satisfy more accurate design, research and monitoring requirements. Prestressing tendons are placed complexly in containment structure, especially near the major equipment hatch area, which is the three-dimensional curve and the friction loss must be considered. Firstly, containment structure and the distribution of prestresseing system are introduced briefly. Furthermore, the calculating process of prestressing tendons losses is put forward in details which take the major equipment hatch as an example. Solid65 element and constitutive model of concrete in ANSYS are introduced in details, and the characteristic of simulating prestressing tendons in ANSYS is analyzed.
Based above study, finite element model of the prestressed structure is built based on some concrete containment by using ANSYS program, the prestressing effect on concrete containment is put forward. The numerical results show that most of the design pressure is beared by the prestressing system under the design-basis pressure condition, and the simulation method is effective with a high accuracy which agrees well with stress tests performed on-site in some operating NPPs such as LINGAO PHASE II.
Research for Crossing Strait Transportation Accident of Nuclear Fuel Assemblies
Authors: Xu Xiaoxiao , Wang Xuexin , Zhang Jiangang , Li Chaoduan , Fan Gao , Tang Rongyao , Cao Lian , Wen Bing 
Published: February 2014

Hainan nuclear power plant (HNPP) is the first nuclear power plant built on China’s Hainan Island. Therefore the nuclear fuel assemblies must transport through the Qiongzhou Strait.
There are two transportation plans to be used in crossing strait transportation of the fuel assemblies. One plan is railway ferry stretching across sea; the other is road vehicular crossing strait on roll-on/roll-off (Ro/Ro) ships.
According to crossing strait transportation scenario and statistical analysis of sea transport accidents in Qiongzhou Strait, three ferrying transportation accidents are considered in this paper. Through research of ship-to-ship collision, fire and sunk, the following conclusions: Collision, fire or foundered are not caused by the leakage of radioactive material, the environmental impact is very small. The accident hazards of crossing strait transportation does not lie in the radiological consequences, but in the effects of public psychology and international repercussions.
Analysis of Dynamic Effects Associated With Postulated Breaks of Piping
Authors: Qin Manqing , Li Xinghua , Zhao Jicheng 
Published: February 2014

High energy pipe break hypothesis and engineering practice shall be taken into account and performed during nuclear power plant design due to defense-in-depth concept unless leak-before-break or break preclusion concept is applied. High energy pipe breaks potentially occur in the high stress area and break locations can be recognized by specific regulations based on mechanical analysis of piping systems. Engineering measures shall be applied to mitigate the consequences caused by high energy pipe breaks later.
From engineering point of view High Energy Line Break analysis requires an essential knowledge in the different fields, such as: structural and dynamic mechanics, hydrodynamics, plant arrangements and so on.
Aims of this paper consist of two aspects: definition of break locations associated with engineering practice of piping stress analysis and a conservative model of anti-whip analysis based on some assumptions. Comparison results are given after analytical investigation of the piping whip analysis with different numerical modeling and tools at the same time. An example of such comparison between transient model and energy balance methods is performed based on some basic assumptions in relation to specific piping layout.
The simplified calculation model appears well for a conservative estimation of pipe whip characteristics, both for maximum displacements of piping and maximum loads impacted on anti-whip restraints.
Incandescent Particle Characterization Emitted by a Cut-Off Grinder During Decommissioning Operations for Evaluating Filters Degradation
Authors: Marchal Pierre , Porcheron Emmanuel , Grehan Gérard , Lafanechere Luc , Walter Jonathan 
Published: February 2014

The use of mechanical or thermal cutting tools in decommissioning of nuclear facilities generates a lot of incandescent particles. These particles may represent a risk of deterioration of the containment barriers associated with a potential risk of fire starting.
The aim of the study is to characterize the incandescent particles emitted by a wheel grinder (in terms of temperature, diameter and velocity) and to follow those parameters all along their path from their emission to their impact on the air filter. The characteristics of particles correlated with a possible loss of filter efficiency should identify the destructing particles for the filter. All the measurement techniques used to characterize the incandescent particles are presented in this article. Particles are characterized in terms of aerodynamic diameter with the help of an Aerodynamic Particle Sizer (APS). The particle velocity is measured with laser velocimetry techniques. An adaptation of a commercial pyrometer is achieved to measure the inflight particles temperature. All of these techniques are being implemented on an experimental facility which reproduces representative conditions of the cutting processes realized during dismantling operations.
Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment
Authors: Brumovsky Milan 
Published: February 2014

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important.
Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing.
The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach.
Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 9 — Effect of Structure Vibration on Rising Bubble Behavior
Authors: Arai Rie , Mizuno Kousuke , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki 
Published: February 2014

Earthquake is one of the most serious phenomena for safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void faction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In case of the earthquake, the fluctuation is not only the flow rate, but also body forth on the two-phase flow and shear forth through the pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.
Therefore, in this research projects, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate effects of vibration on bubbly flow in the components and construct experimental database for validation, we performed visualization experiments of vertical bubbly flow in rectangular water tank on which sine wave vibration was applied. In this paper, results of preliminary visualization experiments were shown including data of positions of bubbles, shapes of bubbles and velocity of bubbles evaluated by the visualization techniques. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble inclination angle was almost same as the cycle of oscillation table velocity.
Preliminary Fuel Cycle Analysis of a Fusion-Driven Subcritical Reactor
Authors: Mao Lisheng , Wang Minghuang , Fu Xuewei , Jiang Jieqiong , Wu Yican 
Published: February 2014

The levelized cost of electricity (LCOE) has been performed to compare two fuel cycle scenarios: a once-through cycle (LWR OT) and a fusion Cdriven reactor, namely FDS-SFB, recycling employing PUREX (Purex-SFB). In order to estimate the LCOEs, the mass flows based on an equilibrium mode were analysed. The sensitivity of the results to variations in key parameters was also performed. A simple dynamic model was also constructed to consider other important factors that characterize a fuel cycle, e.g. resource utilization, environmental effects.
The results of economics are as flows: LWR OT 29mills/KWh, Purex-SFB 48.19mills/KWh. It was found that the capital cost accounts for the largest proportion of the LCOEs. The fuel cycle cost analysis indicates that the FDS-SFB fuel cycle cost will be competitive with the once-through fuel cycle. Also, sensitivity analysis indicates that fuel cycle cost of LWR would be higher than that of the LWR-SFB fuel cycle with the uranium Price rising. Dynamic model analysis indicates that Purex-SFB could reduce the amounts of MA and the amounts of natural uranium considerably compared with LWR OT.
A Study and Development of Primary Circuit Simulation Code for China Experimental Fast Reactor
Authors: Wang Jin , Zhang Donghui , Hu Wenjun , Ren Lixia 
Published: February 2014

A fast reactor is one of recommended candidates of Generation IV nuclear energy systems, which would meet wide requirements such as sustainability, safety and economics for nuclear energy development. To be the China’s first fast reactor, China Experimental Fast Reactor (CEFR) typical technical options are following: 65 MW thermal power and 20 MW electric power, three circuits of sodium-sodium-water, integrated pool type structure for the primary circuit.
To establish modular simulation system for sodium fast reactor, the code which simulated the thermal-hydraulic behavior of primary circuit was developed. The physical models include reactor core, reactor vessel cooling channel, pumps, protection vessel, intermediate heat exchangers, ionization chamber cooling channel, cold sodium pool, hot sodium pool, inlet plenum, and pipes, etc. The code could compute coolant pressures, flow rates, and temperatures in the primary circuit. This module was designed for analysis of a wide range of transients. Although based on CEFR, it can treat an arbitrary arrangement of components.
Neutron Absorbers in CANDU Natural Uranium Fuel Bundles to Improve Operating Margins
Authors: Chan Paul K. , Paquette Stéphane , Bonin Hugues W. , French Corey , Pant Aniket 
Published: February 2014

Safety margins are particularly tight in natural uranium-fuelled CANDU reactors which are refueled on-power. During on-power refueling, the insertion of xenon-free fresh fuel bundles into the reactor core affects the reactor’s excess reactivity in such a way that this could lead to temporary power derating. It is desirable from a fuel management perspective, and to maintain safety margins to eliminate this xenon-free effect and any other power ripples such as the subsequent plutonium reactivity peak. A redesign of the CANDU NU fuel bundle with an appropriate combination of elements, with some including neutron-absorbers, could well address the issue of the xenon-free initial portion of the bundle’s irradiation and also lower the plutonium-peak that occurs shortly thereafter. This may improve the fuel utilization (by further optimizing the fuelling strategy) and provide improved safety margins (by lowering the maximum channel and bundle powers).
The use of neutron-absorbers in fuel design and manufacturing has been a regular practice in Light Water Reactor fuels for more than three decades. In CANDU applications, neutron absorbers have also been considered for the conceptual Advanced CANDU Reactor and the Low Void Reactivity fuel designs, for which the fissile content is made of low enriched uranium (LEU) or MOX fuels. The application to CANDU natural uranium (NU) fuel, however, especially as burnable poisons, is a relative novel approach. The reason for this is that the neutron economy in natural uranium-fuelled CANDU reactors is a prime concern, thus the addition of extra neutron absorbers is generally shunned. In our proposed application of burnable poisons to existing CANDU NU fuel design, because of low excess reactivity for NU fuel, the amount of neutron-absorber is expected to be restricted to small quantities and in a manner whereby the poison effect is restricted to the initial period of excess reactivity of a newly inserted fuel bundle. This implies that the impact on neutron economy would be relatively minimal, but the fuel performance would be significantly improved.
Small amounts and appropriate mixtures of neutron absorbers were selected (approximately 500 mg of absorbers in a CANDU fuel bundle having a nominal weight of 24 kg). Preliminary results indicate that the fuelling transient and the subsequent reactivity peak can be lowered to improve the reactor’s operating margins. A parametric study using the Los Alamos National Laboratories’ MCNP 5 and Atomic Energy of Canada Limited’s WIMS-AECL 3.1 codes is presented in this paper. Details of this project and future work are also to be discussed.