IRREGULAR WARFARE IN THE HIGH NORTH

The strategic importance of the High North––defined in this thesis as the Arctic areas of Finland, Norway, and Sweden––is increasing. Previous research on the region focuses on assessing the often-competing interests of different nations in the High North. From a military perspective, with an emphasis on irregular warfare, this study considers how an irregular warfare approach to defending against a Russian offensive in the High North might unfold. Using data collected from various sources, the study adopts a heuristic methodology. Sources credibility has been verified by thorough searches for conflicting information and differing points of view. The analysis is conducted using historical, environmental, and doctrinal data. Based on the analysis, the thesis addresses the question: What is the most effective way to deter, or if deterrence fails, to counter a Russian offensive using irregular warfare supported by conventional capabilities in the High North? As Russian offensive doctrine is best defeated by irregular approach under some circumstances, an irregular defense and counterattack can provide an effective response if Russia chooses an irregular approach. The study finds that irregular warfare conducted by small, maneuverable units provides an effective solution to the challenge of campaigning against an aggressor in the High North. The thesis recommends developing requirements for logistics, manpower, and command and control for irregular warfare in the High North.Majuri, Finnish ArmyApproved for public release. Distribution is unlimited

An interatomic potential for W-N interactions

N-2 gas is routinely used as a seeding species in fusion plasma to control the amount of power emitted from the plasma by radiation to the tungsten walls of an ITER-like divertor. Nitrogen atoms interact with the plasma-facing materials beryllium and tungsten, and form chemical bonds with the wall surfaces, as well as with plasma hydrogen isotopes, thus raising a special interest in W-N and N-H interactions in the fusion community. In this work we describe the development of an analytical interatomic potential for W-N interactions and benchmark the potential against DFT calculation results for N defects in tungsten.Peer reviewe

Costs and benefits of pig feed salmonella control programme in Finland

201

Deposition in the inner and outer corners of the JET divertor with carbon wall and metallic ITER-like wall

Rotating collectors and quartz microbalances (QMBs) are used in JET to provide time-dependent measurements of erosion and deposition. Rotation of collector discs behind apertures allows recording of the long term evolution of deposition. QMBs measure mass change via the frequency deviations of vibrating quartz crystals. These diagnostics are used to investigate erosion/deposition during JET-C carbon operation and JET-ILW (ITER-like wall) beryllium/ tungsten operation. A simple geometrical model utilising experimental data is used to model the time-dependent collector deposition profiles, demonstrating good qualitative agreement with experimental results. Overall, the JET-ILW collector deposition is reduced by an order of magnitude relative to JET-C, with beryllium replacing carbon as the dominant deposit. However, contrary to JET-C, in JET-ILW there is more deposition on the outer collector than the inner. This reversal of deposition asymmetry is investigated using an analysis of QMB data and is attributed to the different chemical properties of carbon and beryllium.EURATOM 63305

Long-term fuel retention and release in JET ITER-Like Wall at ITER-relevant baking temperatures

The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 °C, respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87% were observed with deposit thicknesses of 10 and 40 μm, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90%. TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experimentsEURATOM 63305

The influence of carbon impurities on the formation of loops in tungsten irradiated with self-ions

The microstructure changes taking place in W under irradiation are governed by many factors, amongst which C impurities and their interactions with self-interstitial atoms (SIA). In this work, we specifically study this effect by conducting a dedicated 2-MeV self-ions irradiation experiment, at room temperature. Samples were irradiated up to 0.02, 0.15 and 1.2 dpa. Transmission electron microscopy (TEM) expectedly revealed a large density of SIA loops at all these doses. Surprisingly, however, the loop number density increased in a non-monotonous manner with the received dose. Performing chemical analysis with secondary ion spectroscopy measurements (SIMS), we find that our samples were likely contaminated by C injection during the irradiation. Employing an object kinetic Monte Carlo (OKMC) model for microstructure evolution, we demonstrate that the C injection is the likely factor explaining the evolution of loops number density. Our findings highlight the importance of the well-known issue of C injection during ion irradiation experiments, and demonstrate how OKMC models can help to rationalize this effect.Peer reviewe

Deuterium retention on the tungsten-coated divertor tiles of JET ITER-like wall in 2015-2016 campaign

Tungsten-coated divertor files exposed during the third JET ITER-Like Wall (ILW) campaign in 2015-2016 (ILW-3) were studied with Secondary Ion Mass Spectrometry (SIMS). ILW-3 campaign contained more high-power plasma discharges and longer plasma time than the earlier ILW campaigns. Measurements showed increased beryllium (Be) deposition on the upper inner divertor, whereas on the outer divertor, Be deposition was lower than during the second campaign in 2013-2014 (ILW-2). Increased intensifies of nickel, molybdenum and tungsten were observed at the surface layer of the inner divertor Be dominated deposits. These layers are probably formed during the high-power plasma discharge phase near the end of the ILW-3 campaign. Compared to the earlier campaigns, D retention on the upper inner divertor was observed to on a similar level than after ILW-2, whereas at the lower inner divertor and most parts of the outer divertor, D retention was lower for ILW-3 than ILW-2. D retention was increased at lower part of outer divertor Tile 7, where Be deposition was slightly increased. Probable reason for the reduction is the higher surface temperature of the files due to higher powers used.Peer reviewe

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

Proceedings of the 22nd International Conference on Plasma Surface Interactions 2016, 22nd PSIAs a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.Peer reviewe

Investigation of deuterium trapping and release in the JET divertor during the third ILW campaign using TDS

Selected set of samples from JET ITER-Like Wall (JET-ILW) divertor tiles exposed in 2015–2016 has been analysed using Thermal Desorption Spectrometry (TDS). The deuterium (D) amounts obtained with TDS were compared with Nuclear Reaction Analysis (NRA). The highest amount of D was found on the top part of inner divertor which has regions with the thickest deposited layers as for divertor tiles removed in 2014. This area resides deep in the scrape-off layer and plasma configurations for the second (ILW-2, 2013–2014) and the third (ILW-3, 2015–2016) JET-ILW campaigns were similar. Agreement between TDS and NRA is good on the apron of Tile 1 and on the upper vertical region whereas on the lower vertical region of Tile 1 the NRA results are clearly smaller than the TDS results. Inner divertor Tile 3 has somewhat less D than Tiles 0 and 1, and the D amount decreases towards the lower part of the tile. The D retention at the divertor inner and outer corner regions is not symmetric as there is more D retention poloidally at the inner than at the outer divertor corner. In most cases the TDS spectra for the ILW-3 samples are different from the corresponding ILW-2 spectra because HD and D2 release occurs at higher temperatures than from the ILW-2 samples indicating that the low energy traps have been emptied during the plasma operations and that D is either in the energetically deep traps or located deeper in the sample.EURATOM 63305

Thermal desorption spectrometry of beryllium plasma facing tiles exposed in the JET tokamak

Corrigendum: 10.1016/j.fusengdes.2018.08.007.The phenomena of retention and de-trapping of deuterium (D) and tritium (T) in plasma facing components (PFC) and supporting structures must be understood in order to limit or control total T inventory in larger future fusion devices such as ITER, DEMO and commercial machines. The goal of this paper is to present details of the thermal desorption spectrometry (TDS) system applied in total fuel retention assessment of PFC at the Joint European Torus (JET). Examples of TDS results from beryllium (Be) wall tile samples exposed to JET plasma in PFC configuration mirroring the planned ITER PFC is shown for the first time. The method for quantifying D by comparison of results from a sample of known D content was confirmed acceptable. The D inventory calculations obtained from Ion Beam Analysis (IBA) and TDS agree well within an error associated with the extrapolation from very few data points to a large surface area.Peer reviewe