bluuebunny/crossref_metadata_2025_split huggingface dataset:
Post-Fukushima Research and Development Strategies and Priorities for Water Cooled Reactor Technology Development
Authors: Yamada Katsumi , Amri Abdallah , Bevington Lyndon , Vincze Pal 
Published: October 2016

The Great East Japan Earthquake and the subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power plants (NPPs) on the north-east coast of Japan and developed into a severe accident at the Fukushima Daiichi NPP, which highlighted a number of nuclear safety issues.
After the Fukushima Daiichi accident, new research and development (R&D) activities have been undertaken by many countries and international organizations relating to severe accidents at NPPs. The IAEA held, in cooperation with the OECD/NEA, the International Experts’ Meeting (IEM) on “Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant” at IAEA Headquarters in Vienna, Austria, 16–20 February 2015. The objective of the IEM was to facilitate the exchange of information on these R&D activities and to further strengthen international collaboration among Member States and international organizations.
One of the main conclusions of the IEM was that the Fukushima Daiichi accident had not identified completely new phenomena to be addressed, but that the existing strategies and priorities for R&D should be reconsidered. Significant R&D activities had been already performed regarding severe accidents of water cooled reactors (WCRs) before the accident, and the information was very useful for predicting and understanding the accident progression. However, the Fukushima Daiichi accident highlighted several challenges that should be addressed by reconsidering R&D strategies and priorities.
Following this IEM, the IAEA invited several consultants to IAEA Headquarters, Vienna, Austria, 12–14 May 2015, and held a meeting in order to discuss proposals on possible IAEA activities to facilitate international R&D collaboration in relation to severe accidents and how to effectively disseminate the information obtained at the IEM.
The IAEA also held Technical Meeting (TM) on “Post-Fukushima Research and Development Strategies and Priorities” at IAEA Headquarters, Vienna, Austria, 15–18 December 2015. The objective of this meeting was to provide a platform for experts from Member States and international organizations to exchange perspectives and information on strategies and priorities for R&D regarding the Fukushima Daiichi accident and severe accidents in general. The experts discussed R&D topic areas that need further attention and the benefits of possible international cooperation.
This paper discusses lessons learned from the Fukushima Daiichi accident based on the presentations and discussions at the meetings mentioned above, and identifies the needs for further R&D activities to develop WCR technologies to cope with Fukushima Daiichi-type accidents.
Radiation-Induced Electrolytic Corrosion of LWRS: (Part 2) — Verification of In- and Out-Core Redox Potential Differences
Authors: Saji Genn 
Published: October 2016

The author recently identified that there should exist a “differential radiation cell” mechanism in the reactor water, prompting “radiation-induced electrolytic (RIE)” phenomena. This mechanism was identified while trying to theoretically reconstruct the potential differences observed in two in-pile test loops; NRI-Rez in Czech Republic and INCA Loop in Sweden.
Part 2 of this series focuses on the theoretical reconstruction of the observed potential differences. Assuming a state of equilibrium, the author tried to develop a formalism by extending the Nernst equation to reproduce the observed redox potential differences. The radiological potential shift term is separated from the Nernst equation where the latter deals only with stable molecular and ionic species. The radiological effect is described as a perturbation term to the Nernst equation representing a potential shift due to radiation-chemical reactions which should diminish to zero without radiation.
The theory generally reproduced the experimental results after fitting the theoretical curve at a single point of the potential for both PWR and BWR-NWC water chemistry environments. This discrepancy is likely due to the “conductive-dielectric property” of the reactor water.
Methodology for Fragility Evaluation of a Seawall Against Tsunami Effects: Part 1 — Overflow and Physical Damage Associated With Tsunami Wave Pressure
Authors: Kaida Hideki , Miyagawa Yoshinori , Kihara Naoto 
Published: October 2016

Successive safety improvements for nuclear power plants (NPPs) have been required by society as well as by regulatory agencies because of the nuclear accident at the Fukushima Daiichi Nuclear Power Plant due to the Great East Japan Earthquake and Tsunami. The establishment of a methodology for the fragility evaluation of seawalls is essential for developing a probabilistic risk assessment (PRA) for tsunamis that is applicable to NPPs where the hazard level of tsunamis is high.
In the present study, fragility evaluation methods of reinforced concrete (RC) seawalls are documented. Two main damage modes of the seawall, namely overflow and physical damage caused by tsunami wave pressure, were the primary focus. Using the documented fragility evaluation methodology, a conceptual study for evaluating the fragility of a RC seawall against overflow and the impact of tsunami wave pressure is performed, and fragility curves are obtained by considering the following uncertainties: evaluation accuracy of the inundation level and tsunami wave pressure, density of the seawater, compressive strength of concrete, yield strength of reinforcement, and evaluation accuracy of the shear capacity.
Development of Rules on Concrete Containment Vessels for NPP
Authors: Nishizawa Tomohiro , Nishikawa Takao , Takiguchi Katsuki , Sugita Hiroyuki , Hijioka Yasuo , Tsuji Hirokazu 
Published: October 2016

“Rules on Concrete Containment Vessels for Nuclear Power Plants” was first published in 2003. It was revised in 2011 and the latest Edition was published in 2015. Endorsement of this Code is expected in the very near future.
There are two types of reactor containments: steel containment vessels (hereinafter referred to as “SCV”) and concrete containment vessels (hereinafter referred to as “CCV”). The former is addressed by Rules on Design and Construction for Nuclear Power Plants, and the latter by Rules on Concrete Containment Vessels for Nuclear Power Plants, including Reinforced Concrete Containment Vessel (hereinafter referred to as “RCCV”), Pre-stressed Concrete Containment Vessel (hereinafter referred to as “PCCV”), and hybrid containment vessels which are required further safety and reliability.
Development and Testing of Fault-Diagnosis Algorithms for Reactor Plant Systems
Authors: Grelle Austin L. , Park Young S. , Vilim Richard B. 
Published: October 2016

Argonne National Laboratory is further developing fault diagnosis algorithms for use by the operator of a nuclear plant to aid in improved monitoring of overall plant condition and performance. The objective is better management of plant upsets through more timely, informed decisions on control actions with the ultimate goal of improved plant safety, production, and cost management. Integration of these algorithms with visual aids for operators is taking place through a collaboration under the concept of an operator advisory system. This is a software entity whose purpose is to manage and distill the enormous amount of information an operator must process to understand the plant state, particularly in off-normal situations, and how the state trajectory will unfold in time.
The fault diagnosis algorithms were exhaustively tested using computer simulations of twenty different faults introduced into the chemical and volume control system (CVCS) of a pressurized water reactor (PWR). The algorithms are unique in that each new application to a facility requires providing only the piping and instrumentation diagram (PID) and no other plant-specific information; a subject-matter expert is not needed to install and maintain each instance of an application. The testing approach followed accepted procedures for verifying and validating software. It was shown that the code satisfies its functional requirement which is to accept sensor information, identify process variable trends based on this sensor information, and then to return an accurate diagnosis based on chains of rules related to these trends.
The validation and verification exercise made use of GPASS, a one-dimensional systems code, for simulating CVCS operation. Plant components were failed and the code generated the resulting plant response. Parametric studies with respect to the severity of the fault, the richness of the plant sensor set, and the accuracy of sensors were performed as part of the validation exercise.
The background and overview of the software will be presented to give an overview of the approach. Following, the verification and validation effort using the GPASS code for simulation of plant transients including a sensitivity study on important parameters will be presented. Finally, the planned future path will be highlighted.
Enhanced Natural Circulation Modeling in the GNF BWR Core Simulator AETNA02 for ESBWR
Authors: Iwamoto Tatsuya , Miranda Dana , Banfield James , Yang Jake , Hizoum Belgacem 
Published: October 2016

GE Hitachi Nuclear Energy’s (GEH) next evolution of Boiling Water Reactor (BWR) technology is the Economic Simplified Boiling Water Reactor (ESBWR). The ESBWR is a natural circulation reactor which employs numerous passive safety features while simultaneously offering a large power output for a relatively small plant footprint. The ESBWR is characterized with shortened fuel length and a tall, partitioned chimney region above the core to promote natural circulation core flow. The ESBWR is studied with Global Nuclear Fuel (GNF)’s advanced core-simulator AETNA02.
AETNA02 is a static, three-dimensional coupled nuclear-thermal hydraulic computer program representing a BWR core. The new GNF lattice-physics core simulator package LANCR02/AETNA02 consists of the two-dimensional method of characteristics based lattice code, LANCR02, which generates cross sections over a range of plant conditions and passes them to the three neutron energy group coarse mesh nodal diffusion code AETNA02. For the thermal hydraulic solution, AETNA02 includes a model which explicitly solves for the flow in each channel, water rod, and bypass region which accompanies a fuel channel. For the solution of a natural circulation plant like the ESBWR, AETNA02 utilizes the Automatic Plant Thermal Hydraulics (APTH) model which includes models for the core, the chimney region, the separator, the dryer, and the downcomer. AETNA02 iterates on the fuel channel flow and the pressure drop by modeling each chimney partition explicitly with fuel channels and bypasses mapped to it in addition to a bypass mixing model for the chimneys. This is relevant because the peripheral channels (low power) will have a lower void fraction while the central channels (higher power) will have a higher void fraction. Thus, the chimneys connected to these channels will each have a different hydrostatic head.
To verify and validate each of the APTH component models, code-to-code comparisons are performed with the GEH TRACG04 code as well as data comparisons with experiments. The TRACG04 computer program is a best-estimate two-fluid transient code. Code-to-code comparisons of the two different methods (two-fluid model in TRACG04 versus drift flux model in AETNA02) are made. In addition, this study attempts to quantify the impact on the core flow distribution that will affect the calculated thermal margins, including the Critical Power Ratio (CPR). The impact of the multiple chimney (MC) versus the single chimney (SC) model is studied. The multiple chimney partition modeling provides additional detail to the core flow distribution that is not considered in the single chimney model. The results confirm and add confidence that the multiple chimney partition modeling will provide improved accuracy in the ESBWR core design.
Multi-Objective Optimization of Electric Heating Pressurizer in Nuclear Power Plant
Authors: Wang Cheng , Yan Chang-qi , Jian-jun Wang , Li Fei-fei 
Published: October 2016

Electric heating pressurizer in nuclear power plant plays a key role in pressure control and protection of primary loop. Small size pressurizer fulfilling the performance requirements has been receiving an increasing amount of attention because of relatively low cost and easy installation. However, the size of pressurizer is sensitive to the thermal-hydraulic and structure parameters and it can only be determined in the thermal design of pressurizer.
In this work, a thermal-hydraulic model of pressurizer was established. Based on this model, the pressurizer inner diameter, primary loop pressure, reactor outlet coolant temperature and pressurizer spray coefficient were taken as design variables, and the weight and volume of pressurizer were taken as optimization objectives. Sensitivity analysis between the design variables and steady-state minimum water volume, steady-state minimum steam volume and steady-state power change volume was implemented respectively to investigate the influence of the variables on the pressurizer thermal hydraulic performance. Sensitivity analysis between the design variables and optimization objectives was also implemented to study the relation between them. The results show that primary loop pressure and reactor outlet coolant temperature obviously effect the fluctuation volume when a positive/negative fluctuation or a power change happens. The primary loop pressure has a negative correlation with the fluctuation volume, while the correlation between the reactor outlet coolant temperature and fluctuation volume is positive. The pressurizer spray coefficient only works in positive fluctuation because there is no need to spray in other conditions. As a structure parameter, the pressurizer inner diameter has little effect on the fluctuation, but it has obvious influence on the weight and volume of pressurizer just as other design variables. Finally, the multi-objective optimization of pressurizer aiming at minimizing the weight and volume was performed by utilizing a hybrid non-dominated sorting genetic algorithm. The optimization result is illustrated in a 2D graph. It shows that there exists a feasible region that the weight and volume of the pressurizer is as much as 10 percentage smaller compared with the original one, which indicates the validity of multi-objective optimization of pressurizer.
HIPed Hard Facings for Nuclear Applications: Materials, Key Potential Defects and Mitigating Quality Control Measures
Authors: Sulley John , Stewart David 
Published: October 2016

Hard facings are used in a number of different types of components generally to provide improved wear, erosion, and corrosion resistance to the substrate. A typical application is to a valve seat which may be subjected to high wear loads and high flow rates such that a hard wearing surface is required.
Traditionally, a common application method has been to weld deposit the hard facing onto a softer substrate, e.g. the main body material of the component. This can result in poor quality micro-structures, e.g. cast/inhomogeneous structures being created that may not provide the required wear/erosion/corrosion resistance. This can be a particular issue for Nuclear applications where the hard facing is cobalt based. If the facing deteriorates and releases cobalt based debris, the debris (crud), if it becomes activated, can contribute significantly to the overall plant radiation activation level. Also, depending upon the geometry, deposition process, the expansion coefficient difference between the facing and substrate, and the cooling rate, a detrimental tensile residual stress can be generated. This can cause cracking of the facing which, depending upon the in-service loading conditions, could lead to crack propagation into the substrate, potentially threatening the structural integrity of the component; a leak or structural failure could result.
One application method to address the potential poor micro-structure that may be created from weld deposition methods, is to use a Hot Isostatically Pressed (HIPed) hard facing. This is where a wear surface is created from the solid state, rather than using a localised melting route, by the HIP consolidation of powder produced by a gas atomisation process. Homogenous, isotropic, finer grained, and defect free microstructures can be created via this method which generally exhibit improved resistance to wear/erosion/corrosion. Also, where the HIP consolidated facing is HIP bonded to the substrate, a beneficial compressive residual stress is created that can arrest any crack propagation into the substrate if cracking of the facing was to occur.
For Nuclear applications it is imperative the material quality of a hard facing is assured. It is essential that care is taken particularly in the production of the powder to ensure it does not contain undesirable defects/contamination.
This paper presents: the benefits that can be achieved with creating component hard facings using the HIP process, the key defects/contamination that may be present, and the quality control measures recommended to assure the material quality.
Combined Systems Approaches Reveal a Multistage Mode of Action of a Marine Antimicrobial Peptide against Pathogenic Escherichia coli and Its Protective Effect against Bacterial Peritonitis and Endotoxemia
Authors: Wang Xiumin , Teng Da , Mao Ruoyu , Yang Na , Hao Ya , Wang Jianhua 
Published: October 2016

ABSTRACT

A marine arenicin-3 derivative, N4, displayed potent antibacterial activity against Gram-negative bacteria, but its antibacterial mode of action remains elusive. The mechanism of action of N4 against pathogenic
Escherichia coli
was first researched by combined cytological and transcriptomic techniques in this study. The N4 peptide permeabilized the outer membrane within 1 min, disrupted the plasma membrane after 0.5 h, and localized in the cytoplasm within 5 min. Gel retardation and circular dichroism (CD) spectrum analyses demonstrated that N4 bound specifically to DNA and disrupted the DNA conformation from the B type to the C type. N4 inhibited 21.1% of the DNA and 20.6% of the RNA synthesis within 15 min. Several hallmarks of apoptosis-like cell death were exhibited by N4-induced
E. coli
, such as cell cycle arrest in the replication (R) and division(D) phases, reactive oxygen species production, depolarization of the plasma membrane potential, and chromatin condensation within 0.5 h. Deformed cell morphology, disappearance of the plasma membrane, leakage of the contents, and ghost cell formation were demonstrated by transmission electron microscopy, and nearly 100% of the bacteria were killed by N4. A total of 428 to 663 differentially expressed genes are involved in the response to N4, which are associated mainly with membrane biogenesis (53.9% to 56.7%) and DNA binding (13.3% to 14.9%). N4-protected mice that were lethally challenged with lipopolysaccharide (LPS) exhibited reduced levels of interleukin-6 (IL-6), IL-1β, and tumor necrosis factor alpha (TNF-α) in serum and protected the lungs from LPS-induced injury. These data facilitate an enhanced understanding of the mechanisms of marine antimicrobial peptides (AMPs) against Gram-negative bacteria and provide guidelines in developing and applying novel multitarget AMPs in the field of unlimited marine resources as therapeutics.
Efficacy of Doxycycline in the Treatment of Syphilis
Authors: Dai Ting , Qu Rui , Liu Jinfen , Zhou Pingyu , Wang Qianqiu 
Published: October 2016

ABSTRACT
Doxycycline is an alternative antibiotic drug for the treatment of syphilis, but data on its efficacy, especially data on its efficacy against late latent syphilis, are limited. A retrospective study was conducted to evaluate the effectiveness of doxycycline for the treatment of patients with different stages of syphilis. Patients who received doxycycline treatment between June 2011 and June 2014 were involved. The serological response to doxycycline was defined as either a negative toluidine red unheated serum test (TRUST) result or a ≥4-fold decrease in titer at 12 months following the treatment. Univariate and multivariate logistic regression analyses were performed to identify factors associated with the serological response. During the study period, a total of 163 syphilis patients were treated with doxycycline, and 118 patients completed doxycycline treatment and the 12-month follow-up. Among the 118 patients, the serological response rate at 12 months was 100.0% (7/7) in patients with primary syphilis, 96.9% (62/64) in patients with secondary syphilis, 91.3% (21/23) in patients with early latent syphilis, and 79.2% (19/24) in patients with late latent syphilis. The total serological response rates were 92.4% (109/118) for preprotocol (PP) patients and 66.9% (109/163) for all intention-to-treat (ITT) patients. In multivariate analysis, patients who serologically responded at 12 months following treatment were positively associated with a higher baseline TRUST titer and an earlier syphilis stage than nonresponders. Our study showed excellent treatment outcomes in patients with different stages of syphilis. Our data, along with those from other reports, support the usage of doxycycline as a good alternative therapeutic option in the treatment of syphilis.
The Colicin E1 TolC Box: Identification of a Domain Required for Colicin E1 Cytotoxicity and TolC Binding
Authors: Jakes Karen S. 
Published: October 2016

ABSTRACT

Colicins are protein toxins made by
Escherichia coli
to kill related bacteria that compete for scarce resources. All colicins must cross the target cell outer membrane in order to reach their intracellular targets. Normally, the first step in the intoxication process is the tight binding of the colicin to an outer membrane receptor protein via its central receptor-binding domain. It is shown here that for one colicin, E1, that step, although it greatly increases the efficiency of killing, is not absolutely necessary. For colicin E1, the second step, translocation, relies on the outer membrane/transperiplasmic protein TolC. The normal role of TolC in bacteria is as an essential component of a family of tripartite drug and toxin exporters, but for colicin E1, it is essential for its import. Colicin E1 and some N-terminal translocation domain peptides had been shown previously to bind
in vitro
to TolC and occlude channels made by TolC in planar lipid bilayer membranes. Here, a set of increasingly shorter colicin E1 translocation domain peptides was shown to bind to
Escherichia coli
in vivo
and protect them from subsequent challenge by colicin E1. A segment of only 21 residues, the “TolC box,” was thereby defined; that segment is essential for colicin E1 cytotoxicity and for binding of translocation domain peptides to TolC.

IMPORTANCE
The
Escherichia coli
outer membrane/transperiplasmic protein TolC is normally an essential component of the bacterium's tripartite drug and toxin export machinery. The protein toxin colicin E1 instead uses TolC for its import into the cells that it kills, thereby subverting its normal role. Increasingly shorter constructs of the colicin's N-terminal translocation domain were used to define an essential 21-residue segment that is required for both colicin cytotoxicity and for binding of the colicin's translocation domain to bacteria, in order to protect them from subsequent challenge by active colicin E1. Thus, an essential TolC binding sequence of colicin E1 was identified and may ultimately lead to the development of drugs to block the bacterial drug export pathway.