bluuebunny/crossref_metadata_2025_split huggingface dataset:
Epidemiology of ocular surface squamous neoplasia in Africa
Authors: Gichuhi Stephen , Sagoo Mandeep S. , Weiss Helen A. , Burton Matthew J. 
Published: October 2013

AbstractObjectivesTo describe the epidemiology and an aetiological model of ocular surface squamous neoplasia (OSSN) in Africa.MethodsSystematic and non‐systematic review methods were used. Incidence was obtained from the International Agency for Research on Cancer. We searched PubMed, EMBASE, Web of Science and the reference lists of articles retrieved. Meta‐analyses were conducted using a fixed‐effects model for HIV and cigarette smoking and random effects for human papilloma virus (HPV).ResultsThe incidence of OSSN is highest in the Southern Hemisphere (16° South), with the highest age‐standardised rate (ASR) reported from Zimbabwe (3.4 and 3.0 cases/year/100 000 population for males and females, respectively). The mean ASR worldwide is 0.18 and 0.08 cases/year/100 000 among males and females, respectively. The risk increases with exposure to direct daylight (2–4 h, OR = 1.7, 95% CI: 1.2–2.4 and ≥5 h OR = 1.8, 95% CI: 1.1–3.1) and outdoor occupations (OR = 1.7, 95% CI: 1.1–2.6). Meta‐analysis also shows a strong association with HIV (6 studies: OR = 6.17, 95% CI: 4.83–7.89) and HPV (7 studies: OR = 2.64, 95% CI: 1.27–5.49) but not cigarette smoking (2 studies: OR = 1.40, 95% CI: 0.94–2.09). The effect of atopy, xeroderma pigmentosa and vitamin A deficiency is unclear.ConclusionsAfrica has the highest incidence of OSSN in the world, where males and females are equally affected, unlike other continents where male disease predominates. African women probably have increased risk due to their higher prevalence of HIV and HPV infections. As the survival of HIV‐infected people increases, and given no evidence that anti‐retroviral therapy (ART) reduces the risk of OSSN, the incidence of OSSN may increase in coming years.
Influence of Boron Carbide on Core Degradation During Severe Accidents in LWRs
Authors: Steinbrück Martin 
Published: October 2013

Boron carbide (B4C) is widely used as neutron absorbing control rod material in light water reactors (LWRs). It was also applied in all units of the Fukushima Dai-ichi nuclear power plant. Although the melting temperature of B4C is 2450 °C, it initiates local, but significant melt formation in the core at temperatures around 1250 °C due to eutectic interactions with the surrounding steel structures. The B4C containing melt relocates and hence transports material and energy to lower parts of the fuel bundle. It is chemically aggressive and may attack other structure materials. Furthermore, the absorber melt is oxidized by steam very rapidly and thus contributes to the hydrogen source term in the early phase of a severe accident.
After failure of the control rod cladding B4C reacts with the oxidizing atmosphere. This reaction produces CO, CO2, boron oxide and boric acids, as well as significant amount of hydrogen. It is strongly exothermic, thus causing considerable release of energy. No or only insignificant formation of methane was observed in all experiments with boron carbide.
The paper will summarize the current knowledge on boron carbide behavior during severe accidents, and will try, also in the light of the Fukushima accidents, to draw some common conclusions on the behavior of B4C during severe accidents with the main focus on the consequences for core degradation and hydrogen source term.
Kinetic Study of Sodium-Water Reaction Phenomena by Differential Thermal Analysis
Authors: Kikuchi Shin , Seino Hiroshi , Kurihara Akikazu , Ohshima Hiroyuki 
Published: October 2013

In a sodium-cooled fast reactor (SFR), if a heat transfer tube in the steam generator (SG) is failed, high pressurized water vapor blows into the liquid sodium and sodium-water reaction (SWR) takes place. SWR may cause damage to the surface of the neighboring heat transfer tubes by thermal and chemical effects. Therefore, it is important to clearly understand the SWR for safety assessment of the SG.
From recent study, sodium (Na)–sodium hydroxide (NaOH) reaction as secondary surface reaction of the SWR phenomena in a SFR was identified by ab initio method [1]. However, kinetics of this reaction is a still open question. It is important to obtain quantitative rate constant of sodium monoxide (Na2O) generation by Na-NaOH reaction because Na2O may accelerate the corrosive and erosive effects.
Differential thermal analysis (DTA) provides us with the valuable information on the kinetic parameters, including activation energy, pre-exponential factor (frequency factor) and reaction rate constant. Thus, kinetic study of Na–NaOH reaction has been carried out by using DTA technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature and the decomposition temperature of sodium hydride (NaH) were identified from DTA curves. Na, NaOH, and Na2O as major chemical species were observed from the X-ray diffraction (XRD) analysis of the residues after the DTA experiment. It was inferred that Na2O could be generated as a reaction product. Based on the measured reaction temperature, the first-order rate constant of Na2O generation was obtained by the application of the laws of chemical kinetics. From the estimated rate constant, it was found that Na2O generation should be considered during SWR.
The results can be the basis for developing a chemical reaction model used in a multi-dimensional sodium-water reaction code, SERAPHIM, being developed by the Japan Atomic Energy Agency (JAEA) toward the safety assessment of the SG in a SFR.
On Distribution of Local Thermal and Mass Flow Rate Parameters of Two-Phase Coolant in the Bundle of 7 Rods
Authors: Dalipagich V. A. , Mayevskiy E. M. , Razumovskiy V. G. , Pis’mennyi E. N. , Koloskov A. E. , Pioro I. L. 
Published: October 2013

Distribution of thermal and mass flow rate parameters of coolant in the cross-section of 7-rod bundle consisting of the tubes of 6.0-mm outside diameter and 1.0-m heated length for two types of discretization of its cross-section into characteristic cells is experimentally investigated by the method of isokinetic sampling. The range of operation parameters corresponds to the conditions of LOCA accompanied by pressure and flow rate decrease relative to the nominal ones and is as follows: bundle-averaged outlet vapor content varies from −0.6 to 0.2, outlet pressure is 6 and 10 MPa, bundle-averaged mass velocity varies from 300 to 1500 kg/(m2s), inlet temperature varies from deep subcooling to (ts − 20) °C, and heat flux rate reaches 1.0 kJ/kg.
It is confirmed by the experiment that the surface in the gap between the peripheral rods is most probable place of burnout in the bundle. Physical model of two-phase flow in the bundle that considers it as the equivalent annular channel with an eccentricity is proposed. The programs of thermal-hydraulic computation of fuel assembly and their adaptation to the experimental data for the two types of discretization into characteristic cells are performed and comparative assessment of them is made.
The data resulted from the study concerning one-phase coolant could be applicable for a reference estimation of heat transfer in future designs of the fuel bundles.
Simulation of a Reinforced Concrete Slab Punching Impact
Authors: Kähkönen Jukka , Varpasuo Pentti 
Published: October 2013

Reinforced concrete wall subjected to an impact by a hard steel missile with a mass of 47 kg and an impact velocity of 135 m/s was one case study in the IRIS 2010 benchmark exercise in OECD/NEA/CSNI/IAGE framework. The wall had dimensions of 2m × 2m × 0.25m and it was simply supported. The perforation of the missile was expected.
Fortum Power and Heat Ltd. participated in the benchmark. In this paper, we present our modeling and blind prediction of the benchmark case. The test results of the benchmark were released after the predictions were made.
Based on the result comparison, we concluded that our model gave conservative results.
Experimental Investigation of Cavitation in the Circulation Pumps With Lead and Lead-Bismuth Coolants
Authors: Lvov A. V. , Bokov P. A. , Shumilkov A. I. , Novozhilova O. O. , Beznosov A. V. 
Published: October 2013

The purpose of this paper is optimization of methods for calculating of flow parts of impeller pumps for high-temperature lead coolant, taking into account specific cavitation performance of the lead coolant [1].
The article presents the studies of cavitation performance of the high-temperature coolant.
The studies were conducted in the lead coolant medium at 450–550 °C, flow rate up to 25–30 m/s and lead flow pressure from 0 to 5.0 kgf/cm2 (atm), at oxygen content in lead from 10−5–10−4 up to saturation, as well as in the presence of solid phase of lead oxides in the flow and in the experimental loop with such coolant.
With a view to improving the representativeness of the research results, conditions of occurrence and characteristics of gaseous cavitation in the lead coolant flow were determined using three independent methods [2, 3].
The research showed that gas cavitation can take place in lead flows in the vane-type pumps and other components of circulation loops followed by apertures of lead flow discontinuity (bubbles, caverns, etc.), filled up with diluted gas with subsequent disappearance of these apertures of flow discontinuity. Traditional cavitation with formation and subsequent collapse of lead vapors in reactor loops with lead and lead-bismuth coolant is not possible.
Cavitation performance of lead coolant is identified.
Study on Mixing Process of Two Component Gases in a Vertical Fluid Layer: Effectiveness on One-Dimensional Natural Circulation
Authors: Mizuno Hiroki , Takeda Tetsuaki , Funatani Shumpei 
Published: October 2013

This study is to investigate an effect of natural convection or natural circulation on a transport process by molecular diffusion in a stratified fluid layer consisting of two component gases. There are many experiment and analysis regarding natural convection or natural circulation in the vertical slot. However, there are few studies on natural convection or circulation and molecular diffusion in the stratified fluid layer consisting of two component gases. It was confirmed that these phenomena appear when the depressurization accident occurs in the very high temperature reactor (VHTR). Therefore it is important to evaluate the transport and mixing processes during the depressurization accident of the VHTR.
The experiment has been performed regarding the combined phenomena of molecular diffusion and natural convection or circulation in a two parallel vertical slots filled with two component gases. The vertical slot consists of the one side heated wall and the other side cooled wall. The other slot consists of the two cooled walls. The dimension of heated wall is 500mm×200mm and thickness is 3mm. The width of the slot is 20mm and the aspect ratio is 25. Combination of nitrogen(N2)/argon(Ar), neon(Ne)/argon(Ar), helium(He)/nitrogen(N2) and helum(He)/argon(N2) was used as the first step of the experiment of the two component gases. The density change of the gas mixture and the gas temperature distribution in the slots were obtained. The mixing process of the heavier gas from the bottom side of the slot filled with the lighter gas was discussed in this paper.
The experimental results showed that the transport phenomena by the molecular diffusion were influenced by the localized natural convection or circulation of the gas mixture in the stratified fluid layer. From the experimental results, it was found that the mixing process by molecular diffusion was affected significantly by the natural convection or circulation induced by the slight temperature difference between both vertical walls.
Insulation Material Deposition and Distribution in a PWR Fuel Assembly Cluster
Authors: Renger Stefan , Alt Sören , Kästner Wolfgang , Seeliger André , Zacharias Frank 
Published: October 2013

Background of experimental and methodical work is the loss of coolant accident (LOCA) with release of fibrous pipe insulation material. Latest investigations were focused on material deposition and distribution (cross mixing) in the reactor core. Therefore, a 2×2 PWR fuel assembly (FA) cluster was constructed. Four shortened PWR-FA-dummies are provided with separated in- and outlets. Every 16×16 fuel rod dummy consists of 20 control rod simulators, two spacers, FA-head and FA-bottom with a 3.5×3.5 mm integrated debris-screen filter (IDF). The cluster is encased in an acrylic housing for visual observation. It is connected with the test facility “Zittau Flow Tray” (ZFT), a simplified sump model, which allows inclusion and investigation of complex phenomena like material sedimentation in the sump and strainer blockages. A well mixing of air in the fluid was also considered by free jet expansions and flows through full cone-nozzles as well as marginal air entrainments.
This Paper includes descriptions of applied measuring techniques (digital image processing, thrubeam laser sensors etc.) and an overview of all considered boundary conditions. Experimental results, aiming at the development, implementation and verification of multiphase flow and strainer models, are presented.
A Preliminary Study of the MSFR Dynamics
Authors: Guerrieri Claudia , Aufiero Manuele , Cammi Antonio , Fiorina Carlo , Luzzi Lelio 
Published: October 2013

In the last years, increasing interest has been focused on an innovative concept of Molten Salt Reactor (MSR) characterized by a fast neutron spectrum that combines the favorable characteristics of MSRs adopting molten salt fluorides both as fuel and coolant with those ones of fast neutron reactors. As a matter of fact, the Molten Salt Fast Reactor (MSFR) has been recognized as a long term alternative to solid-fuelled fast neutron systems and has been identified as reference Gen-IV configuration. Although considerable studies have been carried out for the analysis of the graphite-moderated MSR dynamics, the adoption of a fast spectrum configuration without graphite in the core is expected to notably modify the dynamic behavior of the system, thus requiring further investigation.
In this paper, a preliminary analysis of the MSFR dynamics is performed allowing for the evaluation of the impact of some safety parameters (e.g., feedback coefficients and delayed neutron fraction) on the system behavior for different fuel cycle strategies. For this purpose, a simplified non-linear one-dimensional model of the primary circuit has been developed and the dynamic response of the system has been investigated with reference to different significant transient initiators, namely: unprotected transient overpower, unprotected loss of flow, and unprotected loss of heat sink.
These analyses are thought to give a basic understanding of the MSFR dynamics, as well as significant indications in terms of the system safety parameters.
Experimental Study of Heat Transfer in a 7-Element Bundle Cooled With Supercritical Freon-12
Authors: Richards G. , Samuel J. , Shelegov A. S. , Kirillov P. L. , Pioro I. L. , Harvel G. 
Published: October 2013

Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are technical difficulties in testing and experimental costs at high pressures, temperatures and heat fluxes. Also, there are only a few SCW experimental setups currently in the world capable of providing data. SuperCritical Water-cooled nuclear Reactors (SCWRs), as one of the six concepts of Generation IV reactors, cannot be designed without such data. Therefore, a preliminary approach uses modeling fluids such as carbon dioxide and refrigerants instead of water are practical. In particularly, experiments in supercritical refrigerant-cooled bundles can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) with the critical pressure of 4.136 MPa and the critical temperature of 111.97°C.
A set of experimental data obtained at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) in a vertically-oriented bundle cooled with supercritical R-12 was analyzed. This dataset consisted of 20 runs. The test section was 7-element bundle installed in a hexagonal flow channel with 3 grid spacers. Data was collected at pressures of approximately 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values of mass flux were ranged from 400 to 1320 kg/m2s and inlet temperatures ranged from 72 to 120°C. The test section consisted of fuel-element simulators that were 9.5 mm in OD with the total heated length of about 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 thermocouples.
Analysis of the data has confirmed that there are three distinct heat-transfer regimes for forced convention in supercritical fluids: 1) Normal heat transfer; 2) Deteriorated heat transfer characterized with higher than expected wall temperatures; and 3) Enhanced heat transfer characterized with lower than expected wall temperatures. It was also confirmed that the effects of spacers are evident which was previously observed in sub-critical experimental data. Further analysis needs to be conducted on the deteriorated heat transfer phenomena for low mass flux cases which represent accident scenarios. This can be done by designing a natural circulation experimental test loop.
Identification of Non-Safety Systems to Withstand Beyond Design Basis Events
Authors: Nygaard James , Wan Ping , Chan Desmond , Barrientos Sara 
Published: October 2013

As an aftermath of the natural disasters affecting the Fukushima Daiichi nuclear power plants in Japan, there has been great attention to provide assurance of safety of nuclear power plants around the world. Accordingly, many countries are requiring “stress tests” for their plants to assess the ability to withstand disaster scenarios for which they were not originally designed. Additional efforts are underway to capture lessons learned related to the operation of critical or major systems. Each operator and each country’s regulatory authority may be at different levels of completion for these activities. However, effects on non-safety related or peripheral systems have not been specifically addressed as standalone items or in an integrated systems approach.
This paper seeks to produce an initial assessment of vulnerable systems, structures or components of non-safety related areas that may become critical to the safe operation of a nuclear plant or to the first steps to maintain stability of the plant during a postulated beyond design basis event. The same assessment is valid for events of significant magnitude, or for events affecting the entire site or region, even if a plant’s design basis is not exceeded.
The initial assessment is based on widespread events, such as at the Fukushima Daiichi station, with focus on large nuclear power reactors. Certain peripheral plant systems support plant operators and staff or emergency responders such as by affording communication or physical access to plant areas. Other peripheral systems support plant operation or recovery, for example provision of diverse power supply or cooling means. Passive components common to multiple systems such as cables and piping are also assessed. Once vulnerable systems, structures or components are identified, various modifications or mitigation approaches will be discussed.
Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs
Authors: Takamatsu Kuniyoshi , Sawa Kazuhiro 
Published: October 2013

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results.
We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.
Supercritical-Water Experimental Setup for In-Pile Operation
Authors: Miletić M. , Růžičková M. , Fukač R. , Pioro I. , Peiman W. 
Published: October 2013

The main goal of the Generation-IV nuclear-energy systems is to address the fundamental research and development issues necessary for establishing the viability of next-generation reactor concepts to meet future needs for clean and reliable energy production. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable.
One of the six Generation-IV concepts is a SuperCritical Water-cooled Reactor (SCWR), which continues the utilization of well-known light-water-reactor technologies. Research Centre Rez Ltd. has taken part in a large European joint-research project dedicated to Generation-IV light-water reactors with objectives to contribute to the fundamental research and development of the SCWRs by designing and building a test facility called “SuperCritical Water Loop (SCWL)”.
The main objective of this loop is to serve as an experimental facility for in-core and out-of-core corrosion studies of structural materials, testing and optimization of suitable water chemistry for future SCWRs, studies of water radiolysis at supercritical conditions and nuclear fuels. This paper summarizes the concept of the SCWL, its design, utilization and first results obtained from non-active tests already performed within the supercritical-water conditions.