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The opportunity for online engagement increases possible exposure to potentially risky behaviors for teens, which may have significant negative consequences. Effective family communication about online safety can help reduce the risky adolescent behavior and limit the consequences after it occurs. Our paper contributes a theory of communication factors that positively influence teen and parent perception of communication about online safety and provides design implications based on those findings. While previous work identified gaps in family communication regarding online safety, our study quantitatively identified the factors that significantly contribute to parents’ and teens’ differing perceptions. We analyzed data from a survey of 215 teen-parent pairs through a cross-sectional design and examined the factors that contribute to increased family communication about online safety. For parents, active mediation, technical monitoring, and a perceived positive affect of the teen were associated with higher levels of family communication. Our results were similar for teens, except that the teen’s online safety concern and parental monitoring were also positively associated with increased family communication, while restrictive mediation was associated with lower levels of family communication. Many existing designs for online safety support a restrictive approach, despite teens not wanting technical restrictions. A key implication of our findings is that teens view active mediation and monitoring positively in respect to family communication. Contrary to mainstream narratives, this finding suggests that teens value parental involvement and do not desire complete independence online. By examining specific mechanisms which can hinder or improve family communication between parents and teens regarding online safety, we recommend solutions that give teens an active role in their online safety and facilitate effective family communication through cooperation between both parties, rather than technologies that promote parental restriction.

Considering the high mobility of tritium through the package in which it is contained, the new 50-year storage concepts proposed by the French Alternative Energies and Atomic Energy Commission (CEA) currently provide a solution adapted to the management of waste with tritium concentrations higher than the accepted limits in the disposals. The 50-year intermediate storage corresponds to 4 tritium radioactive periods i.e., a tritium reduction by a factor 16. This paper details the approach implemented to define the waste acceptance criteria (WAC) for an interim storage facility that not only takes into account the specificity of tritium provided by the reference scheme for the management of tritiated waste in France, but also the producers’ needs, the safety analysis of the facility and Andra’s disposal requirements. This will lead to define a set of waste specifications that describe the generic criteria such as acceptable waste forms, general principles and specific issues, e.g. conditioning, radioactive content, tritium content, waste tracking system, and quality control. This approach is also a way to check in advance, during the design phase of the waste treatment chain, how the future waste could be integrated into the overall waste management routes and identify possible key points that need further investigations (design changes, selection of materials, etc.): the results obtained are shown. It supports the nuclear operator to ensure the compliance of the future waste packages with the disposal acceptance criteria.

The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted.

Fusion machines like the ITER experimental research facility will use tritium as fuel. Therefore, most of the solid radioactive waste will result not only from activation by 14 MeV neutrons, but al...

This paper summarises the present status of the ITER tritiated waste management strategy under development in France. This paper describes the specific challenges posed by this radioactive waste containing tritium as well as the solutions planned for the various waste categories and the implementation expected for the ITER tritiated waste, including the features of the future interim storage facility called INTERMED. Several options to reduce temporary storage duration as well as to minimise out-gassing rates and tritium discharges into the environment are under study, the related issues and the preliminary results obtained are shown. The first lessons learned for fusion development and their extrapolation to future reactors are outlined based on four parameters: materials, operating temperature, fuel cycle efficiency and tritium removal technologies.

On his first day in office, on 1 December 2018, freshman President of Mexico, Andrés Manuel López Obrador (AMLO) delivered a speech outlining 100 policy priorities of his administration. The present study analyzed the contributions of this government’s program relating to food security and their environmental implications, and whether they contributed to strengthen the state or improved human security, considering that the poor and marginalized were at the center of AMLO's campaign. In total 45 policy priorities were geared to consolidate the state, while 55 contributed to improving human security. Only six were related to food security, including stipends to food producers and purchasing grains at guaranteed prices, a fertilizer distribution program and subsidies for cattle husbandry and fisheries/aquaculture. These programs contributed to advancing 10 of the 17 Sustainable Development Goals, especially those related to Zero Hunger and Reduced Inequalities. Various policy programs had explicit considerations towards climate change and land degradation, including the exclusion of natural protected areas from agricultural subsidies, and recognized that food production is vulnerable to climate change. The four agricultural programs analyzed may advance AMLO’s goal of avoiding food imports, while curbing rural poverty. However, available evidence is mixed regarding animal acquisition loans, which are likely to have adverse environmental outcomes. Finally, the program for developing agroforestry operations is already contributing to deforestation, and further ecosystem degradation is most likely to occur from the introduction of timber and fruit species to natural forests as this program does not preclude the inclusion of recently cleared plots. If human development goals are to be reached, along with fulfilling the international commitments on sustainable development and environmental conservation, policies need to be implemented that simultaneously tend to a booming transnational industry, while bringing forward the rural poor, who amount to nearly half of the country's population.

DEMO is the main step foreseen after ITER to demonstrate the technological and commercial viability of a fusion power plant. DEMO R&D requirements are usually identified on the basis of the functions expected from each individual system. An approach based on the analysis of overall plant functional requirements sheds new light on R&D needs. The analysis presented here focuses on two overall functional requirements, efficiency and availability. The results of this analysis are presented here putting emphasis on systems not sufficiently considered up to now, e.g. the heating and current drive systems, while more commonly addressed systems such as tritium breeding blankets are not discussed in detail. It is also concluded that an overall functional analysis should be adopted very early in the DEMO conceptual design studies in order to provide a fully integrated approach, which is an absolute requirement to ensure that the ambitious goals of this device will be ultimately met.

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δBT = 0.08% up to δBT = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, βN values above the ‘no-wall limit’ (βN ∼ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.

Foraging is made up of three steps: first, to search and encounter a suitable foraging patch; next prey has to be found, and finally the prey has to be caught and eaten. Behavioral responses to anthropogenic noise may lead to a disruption of vital activities within this process, such as area avoidance, or reduced abilities to locate or catch prey. Gaining insight into these behavioral effects starts with a thorough understanding of within- and between-individual variation in the baseline behavior of experimental conditions. In this study, we analyzed control trials for two captive harbor porpoises that were tested for spatial behavior in a set-up for experimental sound exposure. Data from trials without any experimental sound exposure were used to investigate relationships between the response variable - time spent away from the preferred area - and two other behavioral metrics: swimming speed and surfacing rate. The results show that these metrics can be used as independent measures, the first being correlated to the response variable, the second uncorrelated. Combining the two makes a better and more complete judgment. Performing an adequate exploration of the variation in behavior during control trials is important, as it will aid scientists in revealing and interpreting effects of disturbance in sound exposure trials.

As part of its mission to prepare the operation of ITER, a major programme of enhancements has just been completed on the JET tokamak. These enhancements include a complete replacement of the plasma-facing components in JET, from carbon-based to the combination of beryllium and tungsten foreseen for ITER, an upgrade of the neutral beam heating available on JET from 20 MW/short pulse to 30 MW/long pulse operation, the installation of a high frequency pellet injection system for plasma fuelling and ELM control studies, an upgrade to the JET vertical stability system and a suite of new diagnostics. The future JET programme is foreseen to proceed progressively from a test of fuel retention in the standard regimes of ITER operation towards more aggressive, high performance experiments that will demonstrate the operating space limits with the new wall. Depending on the results of the earlier experiments, the exploitation of the enhancements is foreseen to be completed with a deuterium-tritium experiment. This would represent the most integrated test of ITER operational scenarios possible before ITER itself. JET is a cooperative programme funded and exploited in collaboration by all of the European fusion laboratories. As such, JET is a test bed for multi-national use of a single fusion facility, as is foreseen for ITER. Opportunities for broadening the participation in JET to other ITER Parties are presently being explored. If these opportunities can be implemented, JET would provide not only an integrated test of ITER regimes of operation but also a demonstration of how ITER will be operated, even to the extent of including significant numbers of the same team who will eventually operate ITER.

This article presents an overview of the new ITER-like wall project in JET. It aims at an optimal use of JET’s unique features: physical size, plasma parameters most closely to ITER and the capability to handle beryllium and tritium, allowing the study of critical questions related to operating within the limits of the ITER wall materials. A full replacement of the first wall materials is planned (beryllium in the main wall and tungsten in the divertor). This should deliver answers to urgent plasma surface interaction questions such as tritium retention and provide operational experience in steady and transient conditions with ITER wall materials under relevant geometry and relevant plasma parameters. In addition, the JET auxiliary heating power will be upgraded to ∼45 MW, allowing access to ITER-relevant disruption and edge localised modes energy loss densities. This will open access to conditions of melt layer formation both on the beryllium first wall and the tungsten divertor.

In preparation of ITER operation the JET programme has been focussed to consolidate the ITER reference scenario, the ELMy H-mode, and furthermore to mature the Advanced Tokamak Scenarios, candidate for steady state operation on ITER. In type-I ELMy H-modes simultaneously high confinement (fH98 � /1) and high densities (n/n GW � /0.85) were achieved in stationary conditions by several techniques: (a) increasing plasma triangularity (close to ITER values), (b) argon seeding and (c) periodic pellet fuelling. Techniques for controlling high-pressure operation (control of neoclassical tearing modes (NTM)) and improving amelioration of transient divertor heat loads due to ELMs have been elaborated. In Advanced Tokamak Scenarios the improvement of LHCD coupling to the plasma led to reversed shear plasmas, providing reliable access to strong Internal Transport barriers (ITBs) at lower auxiliary heating powers. The successful development of real-time feedback schemes allowed ITBs to be controlled in steady state conditions (7.5 s, 27/tE). In particular the capability to control both the pressure and current profiles has been demonstrated. Studies related to in-vessel material migration have been further extended, in particular with a new quartz-micro balance diagnostic, further tile analysis and an experimental campaign in helium. This led to new insights, underlining the importance of carbon erosion in the main chamber and chemical erosion in the inner divertor on JET. Altogether this might enable the control of Tritium co-deposition. Detritiation technologies are being developed which could find application on ITER. Enhancements of the facility are undertaken, in view of further extending the operational range closer to ITER, thereby supporting the optimisation of ITER operating scenarios and of ITER auxiliaries. # 2003 Elsevier B.V. All rights reserved.

A new divertor (MKII-HP) has been designed to be implemented in JET as part of a possible enhancement programme of the JET facility (JET EP). The aim is to handle up to 40 MW of injected power for 10 s with plasma triangularities up to 0.5 while keeping enough flexibility for other scenarios. The divertor is shaped to optimise the wetting fraction without exposing sharp edges or metallic parts and the general design allows for high halo currents.

STUDY DESIGN A biomechanical investigation using indentation tests in human cadaveric lumbar vertebrae. OBJECTIVES To determine the effect of endplate removal on the structural properties across the lower lumbar vertebral bodies (L3-L5). SUMMARY OF BACKGROUND DATA The structural properties of the vertebral endplates exhibit substantial regional variation. In addition, several recent studies of the endplate-implant interface have suggested that the endplate is not a significant structural element. METHODS The bony endplates on the left or right side of seven intact human vertebrae (L3-L5) were removed with a high-speed burr. Indentation tests were performed at standardized test sites on both sides of the endplate using a 3-mm diameter, hemispherical indenter with a test rate of 0.2 mm/s to a depth of 3 mm. The failure load and stiffness at each site were determined from the load-displacement curves. Three-way repeated measures analyses of variance were used to analyze the resulting data for variation in the anterior-posterior and lateral directions, as well as to determine the effect of endplate removal. RESULTS For the intact endplates, both the failure load and stiffness varied significantly across the endplate surfaces (P < 0.0001), with the posterolateral regions being stronger and stiffer than the central regions. With endplate removal, the mean failure load decreased to about 33% of the intact failure load, a significant drop (P = 0.04), and there was a trend toward greater decreases posteriorly (P = 0.05). With endplate removal, the mean stiffness also decreased significantly (P = 0.01), with the greater decreases occurring laterally (P = 0.04). CONCLUSIONS There was a significant effect of endplate removal on the local structural characteristics of the vertebral endplate.

This paper summarizes the recent work on JET in the three areas of heating, current drive and energetic particles. The achievements have extended the possibilities of JET, have a direct connection to ITER operation and provide new and interesting physics. Toroidal rotation profiles of plasmas heated far off axis with little or no refuelling or momentum input are hollow with only small differences on whether the power deposition is located on the low field side or on the high field side. With LH current drive the magnetic shear was varied from slightly positive to negative. The improved coupling (through the use of plasma shaping and CD4) allowed up to 3.4 MW of PLH in internal transport barrier (ITB) plasmas with more than 15 MW of combined NBI and ICRF heating. The q-profile with negative magnetic shear and the ITB could be maintained for the duration of the high heating pulse (8 s). Fast ions have been produced in JET with ICRF to simulate alpha particles: by using third harmonic 4He heating, beam injected 4He at 120 kV were accelerated to energies above 2 MeV, taking advantage of the unique capability of JET to use NBI with 4He and to confine MeV class ions. ICRF heating was used to replicate the dynamics of alpha heating and the control of an equivalent Q = 10 'burn' was simulated.

The large physical size of the JET tokamak, its heating systems and diagnostics, and its capability to operate with full deuterium-tritium (D-T) plasmas, including high-power tritium neutral beam injection (NBI), give it unique possibilities in fast particle research in fusion plasmas. These have already been used to generate significant (2-3 MW level) power in fusion a-particles in the 1997 D-T campaign. Recent JET experiments have concentrated on two important scenarios of relevance to next-step tokamak devices: the ELMy H-mode plasmas and plasmas with strong internal transport barriers (ITBs). The achieved progress will help in preparation for a possible second D-T experiment on JET. Fast particle studies have also been carried out recently using ion cyclotron resonance heating (ICRH)-accelerated particles and external-excitation methods to study Alfven eigenmodes (AEs). Looking towards the future, the capability of JET will be enhanced by upgrades to the NBI system, ICRH system and various diagnostics. Results of the first JET D-T experiment (DTE1) form a basis on which to elaborate a second D-T experiment (DTE2) which could be proposed after these enhancements. The alpha-physics part of this programme would be divided between the investigation of alpha-particle confinement, heating and loss processes in the 'integrated scenarios' (where the discharge is as close as possible to an ITER-relevant scenario), and dedicated 'alpha-physics' experiments, with specially prepared plasmas. In ELMy H-mode plasmas the fusion performance could roach Q(=P-fusion/P-input) of similar to0.33 at the highest combined heating powers, corresponding to similar to 6x10(-4), allowing a test of the margins of TAE stability in quasi-steady-state conditions. The integrated-scenario fast particle programme could concentrate on the instabilities and heating in plasma regimes with strong steady-state ITBs, with expected Q values similar to0.58 and similar to2x10(-3), demonstrating the compatibility of these operating scenarios with alpha-effects. Excitation of TAEs by alpha-particles in the plasma core could also be studied in such integrated scenarios. An issue which will receive attention is the confinement of MeV energy ions in the centre of ITB plasmas with strongly reversed shear, where the low current density in the centre may lead to the alpha-particles entering loss orbits. In preparation for a D-T campaign, studies of triton burn-up in deuterium ITB plasmas will begin in the 2002 experimental campaigns. Special 'afterglow' experiments to measure TAEs after the termination of the (stabilizing) NBI have already been explored in JET deuterium ITB scenarios and would be planned for DTE2. It is intended to develop special versions of ITB plasmas with dominant ion heating which would maximize the sensitivity to degradation of alpha-heating effects.

(2001). Some Notes on Planning and Producing a Radio Magazine Programme. Caribbean Quarterly: Vol. 47, No. 1, pp. 69-79.

Tore Supra started operation 10 years ago, on April 1, 1988. This tokamak, today the third in size world wide, has several features representative of the needs of future fusion power production devices: it is equipped with superconducting toroidal field coils (NbTi cooled with He II at 1.8 K); actively cooled plasma facing components; non-inductive current drive systems and real time feedback systems for plasma control. These features provide unique capabilities for achieving high-performance long-pulse plasma discharges and studying several key physics issues in truly steady-state conditions. Several key experimental achievements could thereby be obtained, such as: (i) steady-state high bootstrap fraction discharges (IBS/Ip>50%, ≥5 s); (ii) long discharges sustained by non-inductive current drive, with enhanced confinement, in particular, 2 min discharges, with up to 280 MJ of energy coupled to the plasma, and fully non-inductive steady-state discharges held for duration of up to 70 s; (iii) highly radiating discharges with Prad/Ptot>60±10% and confinement as good as the ELMy H-Mode selected for ITER. These results could be achieved thanks to a tight integration between physics and technology. In line with these achievements, Tore Supra is now preparing its evolution, with the CIEL project, which aims at renewing the inner components of the machine. This should bring the overall power handling capability up to about 25 MW, with plasma facing components able to sustain power densities in the 10 MW m−2 range. The pulse length capability of the tokamak should reach 1000 s.

Neutral beam injection (NBI) is one of the candidates for plasma heating and current drive in the new generation of large magnetic fusion devices (ITER). In order to produce the required deuterium atom beams with energies of 1 MeV and powers of tens of MW, negative D− ion beams are required. For this purpose, multiampere D− beam production and 1 MeV electrostatic acceleration is being studied at Cadarache. The SINGAP experiment, a 1 MeV 0.1 A D− multisecond beam accelerator facility, has recently started operation. It is equipped with a Pagoda ion source, a multiaperture 60 keV preaccelerator and a 1 MV 120 mA power supply. The particular feature of SINGAP is that the postaccelerator merges the 60 keV beamlets, aiming at accelerating the whole beam to 1 MeV in a single gap. The 1 MV level was obtained in less than 2 weeks, the accumulated voltage on‐time of being ∼22 min. A second test bed MANTIS, is devoted to the development of multiampere D− sources. It is capable of driving discharges with current up ...