bluuebunny/crossref_metadata_2025_split huggingface dataset:
A restorer‐of‐fertility like pentatricopeptide repeat gene directs ribonucleolytic processing within the coding sequence of rps3‐rpl16 and orf240a mitochondrial transcripts in Arabidopsis thaliana
Authors: Arnal Nadège , Quadrado Martine , Simon Matthieu , Mireau Hakim 
Published: February 2014

SummaryThe pentatricopeptide repeat (PPR) proteins represent a large family of RNA‐binding proteins that have many roles in post‐transcriptional RNA processes within plant organelles. Among the PPR proteins that target plant mitochondria, the restorer‐of‐fertility (Rf) proteins are characterized by their inhibitory action on mitochondrion‐localized cytoplasmic male sterility (CMS) genes in various crop species. Close homologs to known Rfs from radish, petunia, and rice can be identified in most higher plant species and these proteins define the recognized subgroup of Rf‐like (RFL) PPR proteins. In this paper we describe the function of the RFL9 gene from Arabidopsis thaliana, and show that it is associated with ribonucleolytic cleavages within the coding sequences of rps3‐rpl16 and orf240a mitochondrial transcripts in the Col‐0 accession. RFL9 therefore represents an Rf‐like PPR gene that has the potential to compromise the function of an essential mitochondrial gene and whose function is also associated with a mitochondrial orf sharing significant homology with a proven CMS‐causing orf. We observe that RFL9 is active in only a few Arabidopsis accessions genetically close to Col‐0, which supports the idea that the genetic fixation of this gene represents a regional event in the recent evolution of Arabidopsis. Additionally, RFL9 counts among the RFL genes that are probably controlled by short regulatory RNAs, and our results provides a potential explanation for such control, which in the case of RFL9 might have evolved to limit its detrimental effect on rps3 expression.
Neoadjuvant and adjuvant chemotherapy combined with anatomical resection of feline injection‐site sarcoma: results in 21 cats
Authors: Bray J. , Polton G. 
Published: February 2014

AbstractThis study assesses the outcome of two combined treatment strategies for the treatment of feline injection‐site sarcoma (FISS). Twenty‐one cats with primary or recurrentFISSreceived 3 cycles of neoadjuvant chemotherapy with epirubicin (25 mg m−2), then an anatomical resection of the entire muscle compartment containing the tumour was performed based on the findings of co‐axial imaging. Cats then received a further 3 cycles of adjuvant chemotherapy. Follow‐up was performed by telephone contact with a median follow‐up time of 1072 days. Three cats (14%) developed local tumour recurrence at days 264, 664 and 1573 after surgery. A median survival time could not be calculated as over 80% of the study population remained alive or were censored due to death from other causes. When compared to historical controls, the results of this study demonstrate superior rates of tumour‐free survival and disease‐free interval.
Codes Comparison and Analysis of Weld Cladding Ultrasonic Testing in Nuclear Power Plants
Authors: Zhong Zhimin , Min Jian , Li Kai 
Published: February 2014

This paper briefly introduces the weld cladding structure, its common defects during the manufacture and operation stage and its application in pressurized water reactor nuclear power plants. Some ultrasonic testing codes or standards for nuclear power plant pressure vessel or piping, such as ASME BPVC volume V & III & XI, Germany KTA 3201.3 and 3201.4 code, France RCC-M and RSE-M code, and Russia code of light water nuclear power plants were discussed.
The difference of those codes and some feed backs have been analyzed and discussed. Furthermore, these works really benefit the compiling of NB/T 20003.2-2010, Non-destructive Testing for Mechanical Components in Nuclear Island of Nuclear Power Plants-Part 2: Ultrasonic Testing, as China building more and more nuclear power plants.
It was concluded that we shall pay more attention to the inspection of cladding, not only at manufacture stage but in operation outage stage. One of important work is periodically updating the inspection standard revision. It was believed that improving the cladding defects inspection reliability and effectiveness is very important to the safety of nuclear power plants operation in China and in the world.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 10 — Numerical Prediction of Velocity Profile Around Bubble Under Accelerating Condition
Authors: Yoshida Hiroyuki , Nagatake Taku , Takase Kazuyuki , Kaneko Akiko , Monji Hideaki , Abe Yutaka 
Published: February 2014

An earthquake is one of the most serious phenomena for the 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 fraction 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 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 a body force on the two-phase flow and shear force 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, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to develop the predictive technology of two-phase flow dynamics under earthquake acceleration, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT was expanded to two-phase flow simulation under earthquake conditions. In this paper, the bubbly flow in a horizontal pipe excited by oscillation acceleration and under the fluctuation of the liquid flow was simulated by using the expanded TPFIT. Predicted time series of velocity profiles around the bubbles and shapes of bubbles were compared with measured results under flow rate fluctuation and structural vibration. Predicted results were almost same as measured results qualitatively. And it was concluded that the expanded TPFIT can be applied to qualitative analysis of bubbly flow under accelerating conditions.
Research on Chinese Public Mental in Post-Fukushima Era
Authors: Min Li 
Published: February 2014

Fukushima nuclear accident aroused large-scale public panic toward nuclear power development. Due to the limit knowledge of nuclear power, some people feel frightened and fearful for nuclear safety, nuclear radiation as well as nuclear accident. As the energy with clean, stable and high-efficiency, nuclear power always takes imperative and irreplaceable role in Chinese energy program. Therefore, public acceptance and basic knowledge towards nuclear power in post-Fukushima era is facing new unprecedented challenge. How to relieve the panic and frightening of the public and recover the confidence of nuclear power safety is gradually becoming the hot issue among public.
This paper makes detailed investigations of current public mental in China toward nuclear power in post-Fukushima era, analyzes the internal and external causes of the panic feeling and further proposes several countermeasures and suggestions accordingly for safe and health development of nuclear power in China.
Heat Sink Issue: A Specific Safety Issue for NPP Site Selection and Evaluation in the Warm Countries With Extreme Climatic Conditions
Authors: Rodionov Andrei , Mattei Jean-Marie 
Published: February 2014

One of the major issues during the site selection and design of a NPP is the heat sink issue, i.e. installation and availability of reliable and efficient recipient of heat (air, water or both of them) in relation in particular with:
• waste the heat removed from the condenser of the turbines during normal operational state,
• waste the heat removed from all of the safety-related and non-safety-related equipment and systems during normal operational state,
• waste the residual heat removed from nuclear reactor fuel (situated in reactor vessel or tubes), the decay heat removed from nuclear spent fuel (situated in spent fuel pool), and the heat removed from all of the safety-related equipment and systems during abnormal operational state and accident conditions.
There are several “newcomer” countries that are situated in warm climate zones. In such countries, the cooling water design could be a challenge with regards to extreme water and air temperatures, air humidity, absence of an important reliable cold water source inside the country, etc. In addition, some specific items related to natural, industrial and human environment should be considered during the site evaluation, as for example, risk of intensive biota development, oil spills, siltation, low water level events, etc.
This requires particular consideration of the heat sink issue for the site regulatory review during the licensing process. The article discusses the methodology for site regulatory review with regards to the heat sink issue and provides some practical advises on safety assessment and acceptance criteria to be considered.
Spent Fuel Pool Accidents in a Nordic BWR
Authors: Könönen Niina E. 
Published: February 2014

Loss-of-pool-cooling accidents at the spent fuel pools in the reactor hall of a Nordic BWR have been studied using the MELCOR 1.8.6 code. Studies were made with several different calculation nodalizations. Other investigated variables were the total decay heat power of fuel assemblies in the pool, the initiator of the accident (loss of pool cooling or loss of coolant from the pool), the LOCA leak elevation, the alignment of the re-flooding injection and the use of a lid on top of the pool.
From the results it was observed that ensuring natural circulation of air in the fuel is essential in preventing fuel damages. In cases where the air flow is prevented (loss of pool cooling, LOCA break elevation above the bottom of fuel assemblies or a lid on top of the pool) the fuel is damaged in nearly all cases. Only with the pool decay power being sufficiently low (2227 kW) the melting of the fuel was prevented.
If water injection to the SFP can be restarted before the fuel temperature reaches the cladding failure criterion (1173 K), the rewetting of fuel assemblies is successful and the fuel temperatures will quickly decrease and the assemblies will remain intact. The re-flooding of already damaged fuel assemblies will also cool the fuel rods and reduce the radioactive releases to the environment. However, it may result in greater hydrogen releases.
Primary Shielding Design for an Optimized Molten Salt Reactor
Authors: Zhang Zhihong , Xia Xiaobin , Wang Jianhua , Li Changyuan 
Published: February 2014

Molten salt reactor (MSR) system, a candidate of the Generation IV reactors, has inherent safety, on-line refueling and good neutron economy as typical advantages.
An optimized MSR is developed by changing the size of fuel channel and the graphite-to-molten salt volume radio, based on the Molten-Salt Reactor Experiment (MSRE), which was originally developed at the Oak Ridge National Laboratory (ORNL). In this paper, shielding calculations for the optimized MSR are presented. The goal of this study is to determine the necessary shielding to decrease the neutron and gamma dose rate to the acceptable level according to national regulations. The operating temperature of the optimized MSR is designed in the range of 500 °C–700 °C, heat removal is also considered in the shielding design.
The shielding calculations are carried out by using Monte Carlo method. The shielding system of the optimized MSR consists of 7 zones: the core, the core can, the reactor vessel, the thermal shield, the reactor cell containment, the shield tank and the concrete wall. The combinations of shielding materials in the thermal shield were evaluated. The thermal shield filled with carbon steel balls and circulating water gets an excellent shielding performance and heat removing effects. The neutron spectra and dose distributions, as well as the energy deposition over different shields have been analyzed. The total neutron dose rate outside the thermal shield is attenuated by a factor of about 104, and the gamma dose rate by a factor of about 103. These results show that the shielding design could low dose rate to an acceptable level outside the shielding and far below dose limit required.
Investigation of Thermophysical and Nuclear Properties of Prospective Coolants for Generation-IV Nuclear Reactors
Authors: Dragunov Alexey , Saltanov Eugene , Pioro Igor , Ikeda Brian , Miletic Marija , Zvorykina Anastasiia 
Published: February 2014

Recently, a group of countries has initiated an international collaboration, the Generation IV International Forum (GIF), to develop the next-generation nuclear reactors. The GIF program has narrowed the design options of nuclear reactors to the following six concepts:
1) SuperCritical-Water-cooled Reactor (SCWR);
2) Sodium-cooled Fast Reactor (SFR);
3) Lead-cooled Fast Reactor (LFR);
4) Molten Salt Reactor (MSR);
5) Gas-cooled Fast Reactor (GFR); and
6) Very-High-Temperature Reactor (VHTR);
The purpose of this paper is to compare main thermophysical, corrosion, and neutronic properties of the Generation-IV reactors’ coolants within the proposed range of operation.
Performance Analysis of a Novel PXS With Organic Rankine Cycle (ORC) System
Authors: Zhang Shengjun 
Published: February 2014

With the increasing of core thermal power of the nuclear power plant, the decay heat of the core increases in the accident. Therefore, the heat removal capacity of the PXS should be enhanced to fulfill the requirement of core safety.
A new scheme is put forward to improve the cooling capacity of PXS and provides long-term power for station blackout (SBO) accident or loss of normal feedwater. In this system, the Organic Rankine Cycle is incorporated between the hot leg and cold leg of PRHR. The decay heat of the core is the heat source and the cooling pool outside the containment is the cool source. The natural circulation of the primary loop is established due to the density difference. The primary fluid flows into the evaporator of the ORC system, where the working fluid of the ORC system is evaporated. Then the temperature of the primary fluid is decreased. The vaporized working fluid drives the expander, which is coaxially fixed with the fluid pump, to generate the power. Finally, the exhausted vapor flows into the condenser and the residual heat is discharged outside of the containment simultaneously. The working fluid in the condenser is pumped into the evaporator by the fluid pump for liquid supplement and the cycle keeps on working continuously.
A steady state analysis is performed on a 1700MWe nuclear power plant with ORC as the heat removal system. The heat transfer area of the ORC evaporator is fixed as 487.7m2, which is the same as the area of PRHR HX. The efficiencies of fluid pump and expander of ORC system are assumed as 0.75 and 0.8, respectively. The decay heat of the core is about 67.62MWe, which is 1.38% of the core full power. The working fluids are screened and R141b offers excellent performance. The efficiency of fluid pump and expander are assumed as 0.75 and 0.8, respectively. The condensing temperature is assumed as 80°C and the evaporating temperature is 160°C. The results show that 7.83MWe will be generated by the ORC system and the heat transfer area of the condenser is about 994.5m2. The residual heat of 59.79MWe will be discharged to the water tank outside the containment.
Based on the Meta Analysis Thought of Nuclear Power Equipment Maintenance Strategy Optimization
Authors: Li Shuqing , Mou Huaming , Xiao Hongxing 
Published: February 2014

Nuclear power equipment safety and reliable operation has important social meaning and economic significance, especially some scale is huge, and the mechanism is complex, the impact of decommissioning major and the nuclear safety and to protect public health plays an important role in the equipment, to ensure their safety and reliable operation has an important meaning and the higher maintenance demands.
At the same time, the nuclear power equipment maintenance has maintenance field condition multifarious, difficult, short construction period, quality, safety and technology requirements higher characteristic, so to meet maintenance objectives, the nuclear power plant equipment especially the key and important equipment, the urgent need to adopt more advanced, and more scientific strategy for repair.
Meta analysis method is a more popular on the same theme of multiple independent test results are comprehensive statistical method. At present, it is a very effective comprehensive analysis method, and has already developed mature in many domestic and foreign industries and fields.
This paper based on the Meta analysis thought optimization equipment maintenance strategy, working to improve nuclear power equipment especially the key and important equipment maintenance more scientific, reliability and security.
Fracture Mechanic at Design Level for New Reactors
Authors: Faidy Claude 
Published: February 2014

For new plants in particular, or for some replaced components a new failure mode has to be considered: flaw tolerance and rupture.
Different cracks have to be considered at the early stage in design in different locations with different crack sizes and crack locations, in particular for components in the brittle or transition temperature range. In addition, for some location thermal ageing or radiation ageing have to be considered.
Potential fatigue or corrosion crack growth has to be also considered.
Another aspect that needs to consider fracture mechanic is the Leak Before Break justification or the Break Exclusion justification of some components or piping systems.
This paper will summarize how these requirements are considered in French Codes RCC-M, RCC-MRx and RSE-M for normal, upset, emergency, faulted, severe accident and test conditions.
On-going developments, like Appendix G modernization, will be presented to conclude the paper.
Study on SPMO Management of Contractor Special Process Personnel Qualification Under EPC Mode
Authors: Shen Hongwei , Sun Zhiyou , Shen Yang 
Published: February 2014

Under the background of Sanmen AP1000 NPP Project, World’s First AP1000 Project, this article looks into SPMO management over contractor special process personnel qualification under EPC mode.
The project management mode is briefed and the theory basis of the study is discussed, in terms of nuclear safety regulation ASME NQA-1, HAF003 and industrial safety regulations, as well as quality management theory. Contractor special process personnel qualification is generally unmanageable, mainly resulting in negative chain quality problems, objective of the study aims at solving the problems and managerial dilemmas of contractor special process personnel qualification management on site.
Quality assurance system is maintained according to ASME NQA-1, and special process is implemented in compliance with the specification of special process control procedures developed for particular special processes, such as welding, NDE, painting, etc. Personnel training, authorization, certificate, operation experiences and pertinent requirements are specified in personnel qualification procedures. The method and process of special process personnel qualification management is created under SPMO quality assurance system, in the scope of SPMO management, while detailed and comprehensive classification of special process personnel qualification is defined with originality. On the basis of ingenious classification and management practice, the author summarizes a systematic management methodology of contractor special process personnel qualification under EPC mode. Classification management, standardized checklist implementation, and site surveillance are integrated for special process personnel management. Outstanding effect and performance are achieved by demonstrating the systematic management methodology, providing effective experience for follow-up AP1000 project which has significant meaning for enhancing contractor special process control and ensuring AP1000 construction quality.
Architecting “Maintenance Program” for Piping and Pipe Support
Authors: Tsumori Koichi , Fukuhara Yoshizumi , Terunuma Hiroyuki , Yamamoto Koji , Momiyama Satoshi 
Published: February 2014

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations.
The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.
Study on Formation and Development of Surface Wave of Liquid Metal Lithium Jet for IFMIF
Authors: Hoashi Eiji , Yoshihashi-Suzuki Sachiko , Okita Takafumi , Kanemura Takuji , Kondo Hiroo , Yamaoka Nobuo , Horiike Hiroshi 
Published: February 2014

The international fusion materials irradiation facility (IFMIF) presents an intense neutron source to develop fusion reactor materials. The liquid metal Lithium (Li) jet with a free surface is planned as a target irradiated by two deuteron beam to generate intense neutrons and it is thus important to obtain information on the surface wave characteristic for the safety and the efficiency of system in the IFMIF. We have been studying on surface wave characteristics experimentally using the liquid metal Li circulation facility at Osaka University (Li loop) and numerically using computational fluid dynamics (CFD) code, FLUENT.
The CFD simulation has been used in order to establish the mechanism of the formation and development of the surface wave of the liquid Li jet. The introduction of a two-staged contraction nozzle is planned in the IFMIF and the 1/2.5 size of the IFMIF’s nozzle has been also used and tested in our Li loop. These nozzles have a concave wall at each contraction part, and it was then predicted that Görtler vortices in the boundary layer inside the nozzle was generated and flowed out from the nozzle exit at the high velocity condition in our previous simulation. The Li free surface flow simulation including the flow inside the nozzle set in our Li loop was conducted to compare simulation results with experimental results and to evaluate the influence of Görtler vortices on the surface wave formation and development. In our simulation, large eddy simulation and volume of fluid models are used as turbulence model and interface tracking method, respectively. Our simulation result indicates that both transverse vortices due to gas-liquid shear stress and longitudinal vortices induced by Görtler vortices downstream the nozzle exit contribute to the formation of three-dimensional wave of the Li free surface flow at the velocity of 15 m/s. It was found that the vortex structure and the flow pattern under the free surface due to the flow inside the nozzle strongly contributed the development of the surface wave of the liquid Li jet.