Search

The paper presents the results of the efforts concerned with expanding the verification database and estimating the calculation uncertainty of the power density in the steel reflector of lead cooled fast reactor designs based on experiments performed in different years at the BFS critical assemblies by analyzing and revising earlier calculation and experimental studies on the transmission of neutrons through the steel reflector layers. The discussion includes experiments at the BFS-66 critical assembly to model neutron and photon fluxes in the reactor core shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 critical assemblies to model the transmission of neutrons and gamma quanta through the reflector layers of various materials. The information provided in earlier materials with the descriptions of the above experiments has been analyzed and expanded through respective data required to prepare precision calculation models for Monte-Carlo neutronic codes. Precision neutronic models have been developed based on actualized and updated data with a detailed description of the BFS heterogeneous structure and experimental devices, and test calculations have been carried out to confirm their efficiency. The calculations of key neutronic characteristics measured at the BFS-66, -64 and -80-2 assemblies were performed using codes based on the Monte Carlo method (MCU-BR, MCNP, MMK-RF, MMK-ROKOKO) with BNAB-RF and MDBBR50 neutron data and the ROSFOND evaluated neutron data library. The developed precision calculation neutronic models of the experiments discussed can be used to justify lead cooled fast reactor designs, to verify neutronic codes and neutron data, and to evaluate the associated uncertainties.

The article describes the status of development of codes of new generation for the “PRORYV” Project by the end of 2019: twenty-five commercial-grade software products to justify design solutions and safety of power units with fast neutron reactors and liquid metal coolant (sodium and lead) in a closed nuclear fuel cycle. The developed system of codes is multi-physical and multi-scale that allows performing both calculations of the whole installations and high precision calculations of their individual elements. The developed codes offer unique features. Twelve developed codes have already been certified by Rostechnadzor, and six more have been submitted for certification. In addition to creating the software products, a large-scale work is being carried out to conduct experimental studies for code validation that meet modern requirements imposed by the codes: unique measurement techniques have been created; experimental data on flow characteristics of heavy liquid metal coolant (HLMC) in a fuel assembly simulator have been obtained, as well as of “gas-HLMC” interphase interaction after inert gas injection in HLMC and characteristics of heat exchange between the inert gas and HLMC. The results are already used for validation of system and CFD codes used in the “PRORYV” Project.

Analysis of corrosion processes has a major role in justifying the reliability and safety of developed nuclear reactors of a new generation with heavy liquid metal coolants. An approach has been developed which allows practical conclusions to be made with respect to the processes in the given metal-oxide system based on analyzing state diagrams for these systems in the "oxidation potential – temperature" coordinates. The proposed approach relies on a long-term experience of experimental and computational studies concerned with the interaction of various steel grades with molten lead and lead-bismuth, as well as with the transport of metal impurities within these molten metals. The oxidation potential of a metal-oxide system is measured in experimental studies using oxygen activity sensors developed and manufactured at IPPE. The applicability of the proposed approach to analyzing the processes of mass transport in iron-oxygen, lead-oxygen, sodium-oxygen, and iron-water vapor systems has been demonstrated.

The operating experience of Novovoronezh NPP II-1 shows that, in the summer period, the temperature of the cooling water exceeds the design value: this indicates the insufficient performance of the service water supply system. The main factor that has a negative impact on the performance of this system is the formation of carbonate deposits on the cooling tower filler. At Novovoronezh NPP II-1, the cooling tower water distribution system was cleaned from carbonate deposits by the method of combined vibration and aerohydraulic impact. The tested method of cleaning the filler cannot be considered optimal, since the main stage that determines the entire cleaning duration is the assembly/disassembly of the cooling tower filler. It is necessary to continue research on the choice of a strategy for controlling the carbonate deposition rate, taking into account the revealed influence of the design features of the main cooling water pipelines and pipelines of the cooling tower water distribution system on the mechanism of deposit formation in the peripheral spraying area. As compensating measures to ensure the required temperature regime of the turbine plant equipment at Novovoronezh NPP II-1, it is practiced during the summer period to put the standby heat exchangers of the lubrication system and the standby pump of the nonessential services cooling water system into parallel operation. This solution is fraught with the risk of an unplanned decrease in the electrical load if this equipment is turned off in the event of a malfunction. To increase the operating stability of Novovoronezh NPP II-1 and -2 in the summer period, it is proposed to carry out a number of measures aimed at mitigating the negative consequences caused by the elevated service water temperature. Equipment upgrade options are evaluated, e.g., by installing an additional pump for the turbine building services cooling system and (or) laying an additional pipeline to supply part of the makeup water from the Don River directly to the suction pipelines of the pumps of the turbine building services cooling system.

The key papers on estimating the uncertainties in nuclear data deal with the influence of these uncertainties on the effective multiplication factor by introducing the so-called sensitivity factors and only some of these are concerned with the influence of such uncertainties on the life calculation results. On the other hand, the uncertainties in reaction rates, the neutron flux, and other quantities may lead to major distortions in findings, this making it important to be able to determine the influence of uncertainties on the nuclear concentrations of nuclides in their burn-up process. The possibility for the neutron flux and reaction rate uncertainties to propagate to the nuclear concentrations of nuclides obtained as part of burn-up calculations are considered using an example of a MOX-fuel PWR reactor cell. To this end, three burn-up calculation cycles were performed, and the propagation of uncertainties was analyzed. The advantages of the uncertainty estimation method implemented in the VisualBurnOut code consists in that all root-mean-square deviations are obtained as part of one calculation as the statistical method, e.g. GRS (Generation Random Sampled), requires multiple calculations. The VisualBurnOut calculation results for the root-mean-square deviations in nuclear concentrations were verified using a simple model problem. It is shown that there is a complex dependence of the propagation of the root-mean-square deviations in the nuclear concentrations of nuclides in the process of fuel burn-up, and, therefore, further studies need to aim at investigating the influence of uncertainties in nuclear data on the nuclear concentrations of nuclides.

The article presents an experimental confirmation of the operability of neutron concentrators in devices that form and use directed high-intensity thermal neutron beams with elliptical channels made as blocks of profiled graphite and aluminum plates. The effect of neutron reflection from the surface of materials is the basis of a device capable of selecting neutrons by their directions in space. The study experimentally confirmed the efficiency of a moderating-focusing structure (MFS) based on a pack of elliptical neutron mirrors, which makes it possible to form oriented thermal neutron beams from the outgoing neutron flux. To record the effects of selective thermal neutron separation, silicon single-crystal wafers were used, due to which it was possible to obtain portraits of integral neutron fluxes in the reactor. The experiments were carried out in a horizontal experimental channel (HEC-4) at the IRT-T reactor of the National Research Tomsk Polytechnic University. The integral neutron flux was (2.3–3.02)·1017 cm–2. The neutron flux was detected by the change in the specific electrical resistivity of the single-crystal silicon wafers. The effect of concentration of thermal neutrons was recorded both on the block of graphite neutron mirrors and on the block of aluminum thin-walled elliptical mirrors. In the near future, on this basis, it will be possible to solve such problems as extending the reactor life by reducing the hydrogen uptake in the inner walls. In addition, the experiments have proved the possibility of creating anisotropic structures that lie outside the formalism of Liouville’s theorem, in which the surfaces of thermal neutron sinks are formed with subsequent concentration in the areas separated by aluminum or graphite plates.

The paper presents the results of the studies to justify the design solutions for the main circulation pumps of the heavy liquid-metal cooled reactor plant circuits. A substantial difference has been shown in the performance of pumps for the heavy liquid-metal coolant transfer. The studies have confirmed the qualitative difference in the cavitation performance of coolants, the state of the gases and vapors they contain, the influence of supply and discharge devices, and the effects of the impeller blade section performance and geometry and the hub-tip ratio on the pump performance. The studies were performed based on NNSTU’s lead-cooled test facilities with the coolant temperature in a range of 440 to 550 °C and the coolant flow rate of up to 2000 t/h. The outer diameter of the impellers and the straightening devices was about 200 mm, and the thickness of the flat 08Kh18N10T-steel blades was 4.0 mm and that of the airfoil blades was up to 6.0 mm. The pump shaft speed changed in a stepped manner from 600 rpm to 1100 rpm after each 100 rpm. The studies were conducted to justify the engineering and design solutions for pumps as applied to conditions of small and medium plants with fast neutron lead cooled reactors currently under investigation at NNSTU (BRS-GPG). The experimental results can be recommended for use to design other HLMC transfer pumps.

In asymmetric supercapacitors, electric charges are separated and accumulated in conditions of neutron irradiation. The paper presents the results of studying current sources with radioactive isotopes (CSRI) based on supercapacitors of 100 F made by neutron activation of the elements of the substances (BN or SrO) introduced to the electrodes. Following the neutron irradiation with a dose of up to 400 Gy, the supercapacitors turn into CSRIs and have a charging rate of up to 7.8 μV/day. The CSRI charging rate increases in proportion to the activation with a neutron dose up to 14 kGy. This is explained by the production of β-active isotopes as a result of the 14N(n, p)14C and 88Sr(n, γ)89Sr reactions in the supercapacitor electrode materials. Estimation of the radiation-induced charge accumulation rate, with regard for the energy released during β-decay of 14C and 89Sr, excludes the known mechanisms of charge separation in a condensed environment due to the ionization and production of secondary electrons, the production of electron-hole pairs, etc. The electric charge separation and accumulation in CSRIs take place as a result of non-equilibrium thermoelectric phenomena in the near-electrode nanoregions. Nuclear reactions in nanoscale regions produce “temperature gradients” of up to 105 K/nm, this leading to up to 1013 electrons “evaporating” from nanocrystallites at times of ~ 10-11 s.

Increasing fuel burnup is one of the important areas of nuclear power development. Currently, the most common type of light-water reactors is characterized by burnup ratios of about 5%, i.e., only a small fraction of fuel is used to generate electricity. The paper considers the possibility of a significant increase in fuel burnup due by introducing protactinium and neptunium into the fuel composition. The chains of nuclide transformations starting with protactinium and neptunium are characterized by a gradual improvement in the multiplying properties, which ensures increased fuel burnup. At the same time, a situation may be observed when the multiplying properties of a fuel composition are improved during the campaign, which indicates that at a certain point in time the accumulation rate of fissile nuclides from protactinium and neptunium exceeds the accumulation rate of fission products. While protactinium is hardly accessible in sufficient quantities, neptunium is contained in spent nuclear fuel, a significant amount of which is stored in on-site facilities. Therefore, from a practical perspective, the introduction of neptunium into fuel compositions seems to be more preferable. The novelty of the work is the analysis of the effects of protactinium and neptunium on the reactivity coefficients during fuel campaigns. The calculations were carried out for a VVER-1000 type reactor using the SCALE-6.2 software package.

Proton-conducting oxides have potential applications in hydrogen sensors, hydrogen pumps, and other electrochemical devices including the tritium purification and recovery systems of nuclear fusion reactors. Although the distribution of hydrogen (H) in such oxide materials is an important aspect, its precise measurement is difficult. In the present study, the hydrogen solubility and diffusivity behavior of BaZr0.9Y0.1O2.95 (BZY), BaZr0.955Y0.03Co0.015O2.97 (BZYC), and CaZr0.9In0.1O2.95 (CZI) were studied using tritiated heavy water vapor i.e., DTO (~2 kPa, tritium (T) = 0.1%) by converting the photostimulated luminescence (PSL) values of the imaging plate (IP). The samples were exposed to DTO vapor at 673 K for 2 h or at 873 K for 1 h. The disc-shaped oxide specimens (diameter ~7.5 mm; thickness ~2.3 mm; theoretical density (TD) > 98 %) were prepared by conventional powder metallurgy. The IP images of the specimen surfaces of all the three materials T-exposed revealed that BZY showed the most uniform T distribution with the highest tritium activity. The cross-sectional T concentration profiles of the cut specimens showed that T diffused deeper into BZY and BZYC than into CZI. The hydrogen solubility and diffusivity in the CZI specimen were lower than that in the BZY and BZYC specimens. This suggested that barium zirconates were more favorable proton conductors than calcium zirconates., 19 pages, 12 figures

Possible options of organization of two-component energy system with closed nuclear fuel cycle (CNFC) and new business potential associated with provision of CFC services to foreign customers are examined. Dominating role in the development of nuclear power generation is assigned to VVER reactors with gradually increasing fraction of sodium-cooled fast breeder reactors (FR) incorporated within the joint nuclear fuel cycle operated on MOX-fuel. Components of such energy system perform the following functions: 1. Fast reactors: Generate electric power in base-load mode (possibility of fine tuning of reactor power within limited range (100 – 75 – 100%) is assumed); Utilize waste and/or regenerated uranium for re-fueling power reactors, produce plutonium applicable to the maximum extent for manufacturing MOX-fuel for VVER reactors; Incinerate long-lived highly radioactive wastes – minor actinides separated during reprocessing spent nuclear fuel of FR and VVER reactors. 2. VVER reactors: Generate electricity in compliance with manoeuvrability requirements imposed by the utility company operating the energy system; Utilize MOX-fuel instead of UO2 fuel; Are offered for export with the option of returning SNF back to Russia; Plutonium extracted from VVER spent fuel is used for manufacturing MOX-fuel for SFR. 3. Nuclear fuel cycle facilities: Provide reprocessing of SNF from VVER and SFR with extraction of nuclear materials for recycling; Use depleted or reprocessed uranium and plutonium extracted from spent nuclear fuel for manufacturing MOX-fuel; Provide partitioning of RAW for subsequent utilization of minor actinides and reduction of risks of proliferation of nuclear materials, conditioning and disposal of RAW. Russia possesses capacities for establishing the two-component system with CNFC, as well as the new business approach to rendering CNFC services to foreign customers.

The concept of a high power reactor-laser system based on a nuclear pumped optical quantum amplifier (OKUYaN) was formulated at IPPE in the mid-1980-ies. The idea amounted to the use of wide-aperture OKUYaN as an amplifier within the already well-known “master laser – two-pass amplifier with phase conjugation” scheme. The structure of such an amplifier includes a system of two neutron-coupled units – an ignition reactor (RB) and a nuclear pumped laser amplifier (LB). The ignition unit is a compact multi-core pulsed fast neutron reactor. The laser amplifier unit operates on thermal neutrons and, with regard to the neutronics, it is a subcritical booster zone of the ignition reactor unit. Unique reactor-laser complex incorporating demonstration sample of a pulsed reactor-laser system based on OKUYaN (test facility “Stand B”) having no analogues anywhere in the world, was developed and put into operation at IPPE in 1999 for the purpose of substantiation of basic principles of the OKUYaN concept and demonstration of the possibility of its practical implementation, as well as verification of calculation codes and development of relevant equipment elements. Problems overcome in the development and construction of “Stand B” test facility, the choice and justification of the neutronics and laser characteristics of the OKUYaN demonstration sample are discussed in the present paper. Provided are the results of a detailed computational-experimental study of the demonstration sample characteristics, the data from systems studies of direct conversion of nuclear fission energy into laser radiation energy in complex reactor-laser devices and the results of examination of prospects for the development of nuclear-laser power engineering.

The re-modernization of Unit 4 at the Novovoronezh NPP (Novovoronezh-4) made it possible to take a new approach to the problem of extending the VVER-440 reactor plant life and operation. The authors analyze the existing problems of the VVER-440/179 power unit, showing possible solutions to the identified shortcomings and the final state of the updated power unit. Modernization works significantly expanded the range of design-basis accidents from the primary coolant leak from an opening (DN = 100 mm) to the maximum possible, associated with a rupture of the main circulation pipelines (MCP) (DN = 500 mm). A unique experience was gained in using the safety systems of Unit 3, which was finally shutdown for decommissioning, to increase reliability and provide additional redundancy for the safety systems of Unit 4. The results of the performed works showed the correctness of the adopted concept of re-extending the service life of Unit 4 and ensured its compliance with the modern safety requirements in nuclear power engineering, including as it relates to the safety impact of the first-level probabilistic safety analysis model (PSA-1) for internal initiating events.

One of the most important scientific and technical tasks of the nuclear power industry is to assure the reactor equipment life and reliability under random temperature pulsations. High-intensity temperature pulsations appear during the process of mixing non-isothermal coolant flows. Coolant thermal pulsations cause corresponding, sometimes very significant, fluctuations in the temperature stresses of the heat-exchange surface metal, which, added to static loads, can lead to fatigue failure of equipment components. The purpose of this work was to conduct an experimental study of the temperature and stress-strain states of a pipe sample under the influence of local stochastic thermal pulsations caused by the mixed single-phase heat coolant flows. To solve the set problems, an experimental section was created, which made it possible to simulate the process of mixing non-isothermal coolant flows accompanied by significant temperature pulsations. The design of the experimental section allowed us to study the thermohydraulic and life characteristics of pipe samples made of austenite steel (60×5 mm). Some tools were developed for measuring the pipe sample stress-strain state and the coolant flow temperature field in the zone of mixed single-phase media with different temperatures. The measuring tools were equipped with microthermocouples and strain sensors. As a result, we obtained experimental data on temperature pulsations, time-averaged temperature profiles of the coolant flow in the mixing zone as well as statistical and spectral-correlation characteristics of thermal pulsations. Based on the results of measuring the relative strains, the values of fatigue stresses in the mixing zone were calculated. In addition, some devices and methods were elaborated to measure the temperature and stress-strain states of the pipe sample under the influence of local stochastic thermal pulsations. The developed experimental section provided thermal-stress loading of the metal surface at a high level of alternating stress amplitudes causing rapid damage accumulation rates. The results were included in the database to verify the method for assessing the fatigue life of structural materials for nuclear power plants as applied to austenite steel 12Cr18Ni10Ti under the influence of random thermal cyclic loads.

The article presents the results of work aimed at improving of chemical decontamination methods. A brief description of existing chemical decontamination technologies used to remove radioactive contamination (RC) from walls, floors and external surfaces of equipment without dismantling, i.e., in situ, is given. A vector of research aimed at improving the efficiency of fixed RC’s removal is also determined. The first aim of this work is to improve the decontaminating properties of removable polymer coatings used in practice. The following domestic products were chosen as study objects: compositions presented under trademarks VA, VL; and also special formulation ZPS-1M. Modifications of these compounds performed in SPSIT in some cases made it possible to significantly increase the decontamination factors (DF). The best results were obtained for VL compositions: it was found that due to certain additives it is possible to increase the DF for metal surfaces by a factor of 5–35 over the base product. Along with film-forming decontaminating compositions, an alternative patented technique has been developed in SPSIT. The main feature of this technique is usage of sorbent-based composite covering material previously saturated by decontaminating solution. New technique allows to achieve far higher decontamination factors (150–500) when fixed RC is removed from metal surfaces. In addition, it can be applied to polymer and other non-metal materials. One of the main advantages of given technique is a drastic (11–16 times) reduction of time required for carrying out decontamination operations. The obtained results may be useful for further research in this area. Thus, research performed allows to come up with general conclusion: there are possibilities to efficiently remove fixed RC from surfaces using rather simple chemical means. That, in its turn, could be a rational alternative to high-priced robotic decontaminating systems.

The purpose of the study is to develop a model for predicting the process of a critical heat flux state with the VVER reactor core channel steaming. The model describes the dynamics of the nuclear reactor behavior in conditions of uncertainty, which are typical of abnormal situations, based on information on the process of heat exchange in the core process channels. The use of the proposed model leads to an increase in the speed of response due to a simplified procedure to calculate the parameters of the heat exchange process in the reactor core. The quality of the reactor state assessment is improved through the prediction of the heat exchange process parameters and determination of the critical heat flux parameters in the core prior to the onset of surface boiling the potentiality of which is not predicted in modern VVER in-core monitoring systems. A modification of the mathematical model has been proposed which offers the simplest possible way of using the advantages of neural networks in diagnostics. The model can be used to develop systems for diagnostics of in-core anomalies and systems for adaptive control of the VVER-type reactor thermal power.

The purpose of the present paper is the formulation of the analytical version of the resonance coupled-channel model (RCCM) originally developed for D + T → 5He** → α + n nuclear fusion reaction. The integral in the denominator of the Breit-Wigner type is examined in the expression for S-matrix elements of binary processes in this model. Imaginary part of this integral determines the energy-dependent decay width for the near-threshold channel. It is demonstrated that this integral can be calculated explicitly with the Binet representation for the ψ-function (the logarithmic derivation of the gamma function). As the result the explicit expression for the S-matrix elements in the form of analytical functions of the channel momenta are obtained and the equivalence of the RCCM and the effective range approximation (Landau – Smorodinsky – Bethe approximation) is established on this basis. This allows expressing the parameters of the RCCM through the model independent system characteristics: the complex scattering length and the complex effective range. Several sets of model parameters of both approaches that provide a good description of the measured data on D + T → α + n reaction and D-T elastic scattering are derived. By this means we find the location of the S – matrix poles on different Riemann sheets which corresponds to Jπ = (3/2)+ state of 5He and 5Li nuclei. In particular, the location of the resonance (R) and shadow (S) poles is determined: 5He**: ZR = 46.9 – i37.2 (keV) ZS = 81.7 – i3.5 (keV) 5Li**: ZR = 205.7 – i146.8 (keV) ZS = 264.4 + i112.0 (keV). Our results agree well with previous findings. The possible generalizations of the results obtained are discussed.

To license nuclear fuel for nuclear power plants, data on the behavior of fuel elements (FE) under design-basis accidents are required. These data are obtained during tests of fuel assemblies (FA) and single fuel elements in research reactor channels followed by post-test studies in protective chambers. A reactivity-initiated accident (RIA) with an unauthorized release of CPS rods from the reactor core leads to a pulsed channel power increase. This accident can proceed according to two scenarios: without a critical heat flux (CHF) on the fuel element jacket at the final stage and with a dry heat flux. To date, a series of experiments have been carried out according to the first scenario in the MIR reactor channel and the corresponding data on the behavior of fuel elements have been obtained. An urgent task for today is to prepare and conduct reactor experiments according to the second scenario. The main experimental parameter that determines the behavior and final state of the studied fuel elements is their temperature. No experimental data were found on the critical heat flux for the rod bundles in the low coolant mass flow rate region (experiments in the MIR reactor channel can be conducted in the range of 200–250 kg/(m2s)). The available data are in the extrapolation range. The “steady-state crisis” experiment was conducted to obtain data on the critical heat flux value within the specified coolant mass flow rate range in the MIR reactor channel. The test object was a jacket fuel assembly composed of three shortened VVER-1000 fuel rods with a length of 1230 mm (the fuel part length = 1000 mm) installed in a triangular grid at a pitch of 12.75 mm, which is a cell of the VVER-1000 core. This assembly configuration is used for in-pile tests to study the behavior of fuel elements under emergency conditions. The in-pile testing results are presented. The paper shows the possibility of detecting the start and development of a dry heat flux based on the readings of thermocouples located inside the FE kernel. As a result, the directly measured test parameters were used to determine the critical heat flux value.

It is now well-known that under exposure to a helium (He) plasma at elevated temperatures, a tungsten (W) surface is modified by the growth of a fibreform nano-structure referred to as “fuzz”. Formation of W fuzz occurs when a specific set of conditions are met and raises concerns for the behaviour of the ITER W divertor targets. The key open question remains, however, of whether or not fuzz will actually form in ITER.An analysis of existing data shows that the fuzz annealing kinetics can be described by a diffusive-like kinetics law. A growth/annealing equilibrium model is then proposed to take into account Edge Localized Mode (ELM)-induced erosion and the effect of the transient temperature excursion on fuzz growth and annealing. In addition, both the decrease of fuzz erosion rate and thermal conductivity with increasing thickness are treated consistently. The main finding is that as fuzz grows thicker, its thermal conductivity decreases and peak temperature attained during an ELM increases. Given the competition between growth and annealing rates, an equilibrium thickness exists above which fuzz annealing dominates for a given ELM energy density. In the absence of ELMs, the fuzz thickness would increase with the typical t0.5 time dependence. PACS: PSI-23, Keywords: Helium, Tungsten, ITER, Nanostructure, Fuzz

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 D-2 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. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2019.03.012

An effective permeation reduction of gaseous hydrogen isotopes into a metal wall by introducing a barrier is essential in two main fields: prevention of hydrogen-embrittlement in steels and control of the tritium inventory in future nuclear fusion reactors. By far the most important advances and relevant studies originate from the Nuclear Fusion Community where tritium retention is an important issue, influencing safety issues.Stable permeation barriers are searched among materials with the lowest bulk hydrogen solubility and diffusivity. Besides a few specific pure metals, like beryllium and tungsten, dense oxides, nitrides and carbides have been mostly investigated. Coating techniques for preparation of well adhered and perfect barriers are evidently of the same importance as the material selection itself. Most attractive are the techniques where an ad-layer is formed simply by oxidation. Other methods require specific gas environments with strong electric and magnetic fields, which may represent a limit for the ad-layers uniform coverage over large and uneven areas. Evaluation of the achieved barrier performances is another challenging task. Several new methods, which can trace hydrogen isotopes in bulk at very low concentrations, often miss in the determination of their mobility. Also, they do not reveal the role of barrier defects. The classical gas permeation rate method through coated membranes is still the most reliable option to determine the actual Hydrogen Permeation Barrier (HPB) efficiency. At elevated temperature, the hydrogen permeation rate is recorded at the downstream side of a coated membrane exposed to a substantially higher hydrogen upstream pressure. By using modern vacuum instrumentation techniques, even the most effective barriers can be well characterised. Keywords: Hydrogen permeation barrier, Nuclear fusion, Oxides, Nitrides, Carbides

The plasma surface interactions within a fusion device are one of the limiting factors to long pulse, power generating operation. A plasma facing component material will require effective heat tolerance, minimal erosion yield, and minimal fuel retention properties. Tungsten (W) has been selected as the divertor material for the International Thermonuclear Experimental Reactor (ITER) due to its high thermal conductivity and high sputter threshold. However, when W is exposed to high particle flux (>1022 ions/m2s) at high surface temperatures (>600 °C), the surface will develop defects such as pits, blisters, and nano-structured tendrils, reducing the beneficial properties of W. To overcome this limitation, a more radiation tolerant thin film material could be used, such as lithium (Li). In addition, the lithium film can protect the plasma from high-Z W sputtered atoms. In multiple tokamak devices, a Li wall coating, has improved the plasma performance by reducing fuel recycling from the walls, stabilizing the edge plasma and decreasing the number of edge localized modes (ELMs). Since ELMs help eject impurities from the core plasma, the complete suppression of ELMs is detrimental. Methods to regulate the frequency of ELMs have been investigated using gas puffs. In this work we report a new method to control the ELM frequency by tuning fuel recycling via the intrinsic helium (He) ions produced as ash from the deuterium (D) – tritium (T) fusion reaction. In Li films, one mechanism to retain D is via a chemical interaction between Li, O (oxygen), and D. Previous work has shown that when He ions are introduced with D ions, in a dual beam irradiation of Li films on W, a reduction in the dynamic surface D retention is observed. To further investigate this phenomenon, 1–2 um films of Li on W were exposed to sequential irradiations of D and He. The He fluence was ≈5% of the D (3.3 × 1020 ions/m2). The energies for the He and D ions were 1000 eV and 250 eV/amu, respectively and samples were exposed at room temperature. The surface chemistry was characterized with x-ray photoelectron spectroscopy (XPS) to determine changes in retention. The XPS scans were conducted in-situ and in-operando for the irradiations. Our results showed a decrease in the surface retention when He follows D ions and little change in the retention when D follows He. This indicates that He breaks the D retention mechanism in Li. Keywords: Lithium, Tungsten, Deuterium, In-situ XPS, Plasma facing components, Hydrogen retention

A synthetic gas puff imaging (GPI) diagnostic is applied to a gyrokinetic turbulence simulation of Alcator C-Mod with the objective of assessing the impact of turbulent plasma fluctuations on the GPI neutral density field. We find that these effects are significant for much of the run, resulting in qualitatively different light emission patterns relative to the idealized limit of a fixed neutral profile. Spatial and temporal variations in the local neutral density within the GPI gas cloud are shown to be comparable in magnitude to those in the plasma, which would affect the interpretation of plasma fluctuations made using GPI. Assessing the experimental implications of these results will require additional study using validated turbulence simulations. Our findings are compatible with previous theoretical investigations concluding that GPI may unreliably diagnose plasma turbulence more than one neutral mean free path away from the gas source. Although we used a comprehensive, time dependent simulation to arrive at these results, our analysis shows that a frozen-plasma treatment of the neutrals is adequate in this case. That is, we can assume that the neutral transport is too fast to be affected by temporal fluctuations in the turbulent plasma structures. Keywords: Synthetic diagnostic, Gas puff imaging, Neutral transport simulation, Plasma turbulence simulation, Alcator C-Mod

The tungsten divertor of the Experimental Advanced Superconducting Tokamak (EAST) device operates at varying temperatures and strong electromagnetic fields during plasma operation. In order to monitor the thermomechanical and electromagnetic responses of the structural components, a strain sensor system has been proposed to be applied in the EAST device. To develop this technology, two initial experiments on the thermal strain measurement of a tungsten divertor module of the EAST have been carried out using electrical strain gages and optical strain sensors. However, some engineering issues were exposed during the experiments. Under this condition, all these issues are summarized and discussed this time. Moreover, a comparison between the two techniques has been made when applied in the tungsten divertor. To make them more suitable for the application in the tungsten divertor of the EAST, several activities including a thermomechanical simulation, the welding technologies for the sensors’ installation and a thermal stability test on the optical strain sensors were conducted. Based on these progresses, it is expected that the strain monitoring system will be established in the EAST device in the near future. Keywords: Tungsten divertor, Strain monitoring, Electrical strain gage, Fiber Bragg Grating, EAST

The impact of fluid molecules on detachment was studied using the edge fluid code UEDGE for a set of DIII–D L-mode plasmas with varying degree of low-field side (LFS) target detachment. A baseline atom-only, fluid-neutral model was compared to models including diffusive deuterium molecules with user-defined molecular temperatures and molecular temperatures solved via an energy equation, respectively. The UEDGE simulations including molecules with their temperature solved via the molecular energy equation displayed improved qualitative agreement with DIII–D measurements. However, at LFS separatrix electron densities below 1.3 × 10 19 m − 3 the LFS target peak electron temperature was over-estimated compared to the DIII–D measurements by a factor of 1.5–2. Kinetic effects, as the SOL collisionalities for LFS separatrix electron densities below 1.3 × 10 19 m − 3 are ν S O L * ≲ 15 , could contribute to this discrepancy. At LFS separatrix electron densities above 1.5 × 10 19 m − 3 UEDGE predicted LFS target peak ion current and electron densities a factor up to ∼ 4 lower than the DIII–D-measured values for all models. The consideration of molecules in UEDGE simulations induced a stable X-point MARFE, while an X-point MARFE was not achieved with the ion-atom model. The X-point MARFE forms due to a high-density front which connects the high-field side (HFS) and LFS divertor legs over the X-point, and is not observed for the atom-only UEDGE simulations. The high-density front forms due to molecular particle dynamics and cross-field drifts increasing the plasma density towards the X-point in the HFS leg.

Local neutral pressure measurements in the closed small angle slot (SAS) divertor [1,2] on DIII-D show a large increase when the divertor plasma shifts from high recycling into detachment. In-tile pressure gauges were installed to measure the pressure in the near- and far-SOL regions to examine the predicted high neutral pressures and compression. Cross-field drift effects lead to ∼ 10 × higher peak neutral pressure in detachment with the ion ∇B drift toward the divertor compared to out of the divertor, with similar pressures in attached conditions. Drifts also play a role in the neutral distribution in the slot while attached but become less pronounced in detachment once gradient drives reduce. Variation in the outer strike point location found higher neutral pressure and detachment at modestly lower main-plasma density with the strike point positioned away from the designed operation point. Reducing the fraction of the SOL width allowed into the slot increases neutral leakage into the main chamber and increases the main-plasma density required for detachment. Preliminary modeling with the SOLPS code without drifts over-predicts the neutral pressure in detachment by a factor of 2 with the strike point in the designed operation point while more significantly over-predicts the neutral slot compression; experiments show a broader distribution of neutrals through the slot. These measurements are used to help understand detachment and validate divertor design metrics. Keywords: DIII-D, Divertor, Neutral pressure, Detachment

Injection of low-Z granules into high performance discharges on DIII-D has been shown to promptly trigger Edge Localized Modes (ELMs) providing high-Z impurity control without significant plasma degradation. The ability to provide ELM triggering over a range of injection and discharge parameters suggests that the mechanical introduction of granules can be considered as an additional method of impurity control in ITER. Utilizing a spherically symmetric vapor shielding model for granule ablation, benchmarked with impurity granule injections on DIII-D, we simulate the injection of beryllium granules into ITER baseline discharges. By comparing the granule induced ELM triggering size required for deuterium and non-fuel pellets on DIII-D and cross-correlating with a previously simulated JOREK calcuation of D pellet size required for ELM triggering in ITER, we estimate that a beryllium pellet of 1.5 mm diameter should provide reliable ELM triggering on ITER. This size pellet, delivered at 200 m/s should penetrate 3.5 cm past the separatrix, solidly within the H-mode steep gradient region, a location found to be advantageous for ELM triggering with minimal pellet size. Keywords: ELM pacing, granule injection, beryllium

Plasma flow patterns in a super-X-divertor-like configuration are investigated by using a one-dimensional plasma fluid model with the anisotropic-ion pressure (AIP model). The AIP model enables to treat supersonic plasma flows self-consistently by describing the parallel plasma momentum transport with a hyperbolic equation keeping the finite effect of the parallel viscosity. A supersonic plasma flow in the diverging-magnetic-field divertor region is generated due to the magnetic nozzle effect. It is found that the sonic-transition point has bifurcation characteristics. A numerical solution from the AIP model agrees well with an analytical one. The Braginskii’s plasma fluid model, on the other hand, creates various unphysical profiles in the supersonic plasma flow region for different boundary conditions of the plasma flow at the sheath entrance. It is also found that a particle source/sink in front of the target brings about generations of subsonic/supersonic plasma flow profiles. Moving the target position, it is found that a discontinuous change in the sonic-transition point and corresponding profile of plasma flow can happen due to the bifurcating characteristics of the sonic-transition point. Keywords: Super-X divertor, Plasma fluid model, Anisotropic ion pressure, Magnetic nozzle effect, Supersonic plasma flows, Bohm criterion

We present results from atomistic simulations of sputtering and near-surface implantation of concurrent He, D, and T bombardment of cubic silicon carbide (3C-SiC). This is achieved by first establishing a many-body interatomic potential parameter set to treat interactions of He and hydrogenic species in 3C-SiC informed by ab-initio calculations. To obtain sputtering yields we perform both classical molecular dynamics and binary collision approximation simulations for normal incident particles having energies ranging from 25 to 800 eV. We find that due to differences in species surface binding energy of various crystallographic surfaces in 3C-SiC, the sputtering yield of Si is significantly less than that of C, but sputtering yields show limited sensitivity to crystallographic surface orientation. An exception to this occurs when the terminating crystallographic surface plane is more rich in Si rather than C, resulting in comparable sputtering yields of Si and C. The influence of temperature on sputtering is explored and shows limited effect. Finally, the nature of implanted He, D, and T within 3C-SiC surfaces is investigated to understand implantation profiles and stability of defects. Keywords: Sputtering, Silicon carbide, Surface binding energy, Molecular dynamics

ITER-grade tungsten (W) specimens were exposed to similar deuterium (D) plasma condition (ion flux density of 6.0 × 1021 D m − 2s−1, D ion fluence of 5.0 × 1025 D m − 2) at the surface temperature of 623 K. Thermal desorption spectroscopy was used to measure total D retention in one W specimen after D plasma exposure. Glow-discharge optical emission spectroscopy was used to measure D depth profiling from the other W specimen exposed to the similar condition, and deep D trapping up to 25 μm was observed. When the normalized D depth profile was used with a reaction-diffusion code TMAP7 to model experimental D desorption behavior, an excellent agreement to experimental results was achieved. The modeling results suggested that the predominate mechanism of the deep D trapping might be D trapping in intrinsic intergranular cracks in ITER-grade W. Keyword: Tritium retention, Neutron-irradiation, Plasma facing-components

Plasma-facing materials in future large-scale fusion reactors must be designed to withstand high heat fluxes from extreme off-normal events such as edge localized modes and unmitigated plasma disruptions. The erosion rates of possible tungsten-alternative materials are tested under high heat flux conditions at the DIII-D National Fusion Facility. High-purity β-3C CVD silicon carbide was exposed alongside MAX phase ceramic Ti3SiC2 to both L- and H-mode plasma discharges in the DIII-D divertor. Samples survived average heat fluxes ranging from 2–10 MW/m2 over 16 s. A new micro-trench erosion measurement technique was successfully implemented and measured Ti3SiC2 and SiC erosion rates of 0–9 nm/s and 27–73 nm/s, respectively. Additionally, average ion impact angle estimates for an incident B-field angle of ∼1.5° from surface parallel were made using micro-trench impact patterns. Measurements ranged from θ = 24o–34o with respect to Bt and ϕ = 51.5o–55o below the surface normal.

Pure W, and W–0.5wt%ZrC (WZC) and W–2.0wt%Zr–0.5wt%ZrC (WZZC) alloys are fabricated using vacuum hot pressure (VHP) method to investigate the effects of trace ZrC or Zr/ZrC on the mechanical properties and microstructure of tungsten. It was found that from W to WZC and to WZZC, the average grain size was refined from 6.2 µm of W to 5.8 μm of WZC and finally to 4.8 um for WZZC and relative density increase continually from 97.0% to 98.0% and to 98.5%. The tensile tests indicated that ZrC or ZrC/Zr significantly increases the strength and ductility/plasticity. The comparison of VHPed and spark plasma sintered (SPSed) specimen are made. The microstructure analysis indicate that the W–Zr–O and/or Zr–C–O particles resulting from the reaction of Zr or ZrC with free O pin grain boundaries (GBs) of tungsten and inhibits the migration of GBs, and then refining the grain size. The synergistic effects of the decrease of free O impurity and nano-sized ZrC dispersion strengthening strengthen GBs and improve the mechanical properties. Keywords: Tungsten, Trace Zr/ZrC, Mechanical property, Microstructure

Lithium (Li) coating of plasma-facing components (PFC) has led to improved plasma performance in fusion experiments due to the effectiveness of Li in retaining hydrogen (H) isotopes and thus reducing recycling. Since Li readily reacts with background gases present in fusion experiments, it is important to understand and parameterize the time dependence of deuterium (D) retention in Li and Li compounds for applications of Li under future long-pulse plasma conditions. D retention in pure Li and composite LiO and LiCO films was studied as a function of time after these films were exposed to 450 eV D2+ ion irradiation. The amount of deuterium retained in both Li and LiO films at 300 K decreased at the same rate by 45% after 16 h. The amount of D retained in LiCO films was found to be independent of time up to 3 days. Increasing the temperature of a pure Li film to 420 K increased this rate of decrease by 50%, while increasing the thickness of the lithium film from 3 to 16 monolayers had no effect on the decrease rate. Keywords: Deuterium retention, Lithium, Lithium compounds, Time dependence

During its second operational campaign (OP 1.2a) W7-X operated with an edge island divertor. Much higher densities (1020m−3) than in to the previous, limiter campaign were achieved. In order to asses this improved performance and the nature of transport with regard to the expected higher densities and other parameters such as the electron and ion temperature, the electric field and plasma flow, measurements at the plasma edge were conducted with reciprocating probe, mounted on a Manipulator. The probe contains a set of Langmuir pins, Mach probes, compensation coils and a set of 3D pick up coils for the measurement of the local magnetic field and fluctuations, and an ion sensitive probe. It has been observed, in addition to differences due to the magnetic configurations, that the magnetic topology is sensitive to the plasma conditions, especially to changes due to a evolving toroidal plasma current. Measurements from divertor cameras showed that an increasing toroidal current in the standard configuration broadened the heat flux and shifted the strike line considerably. These down-stream observation can be complemented with the measurements at the mid-plane with the manipulator mounted probes. During the experiments the edge island remnant was identified with the measured edge plasma profiles and compared with the predictions from the field line tracing code. Those electron densities and temperatures have been used as an input for understanding edge transport with the EMC3-EIRENE modeling. The results from EMC3-EIRENE also help improve the caculated density profiles, since the modelling provides a profile of the effective charge Zeff. Keywords: W7-X, Langmuir probes, Magnetic islands, Plasma edge transport

A micro-structuring of the tungsten plasma-facing surface can strongly reduce near surface thermal stresses induced by ELM heat fluxes. This approach has been confirmed by numerical simulations with the help of ANSYS software. For experimental tests, two 10 × 10 mm2 samples of micro-structured tungsten were manufactured. These consisted of 2000 and 5000 vertically packed tungsten fibres with dimensions of Ø240 µm × 2.4 mm and Ø150 µm × 2.4 mm, respectively. The 1.2 mm bottom parts of the fibres are embedded in a copper matrix. The top parts of the fibres have gaps about of 10 µm so they are not touching each others. The top of all tungsten fibres was electro-polished. A Nd:YAG laser with a pulse duration 1 ms and a pulse repetition frequency of 25 Hz was used to simulate up to 105 ELM-like heat pulses. No damage on either of the micro-structured tungsten samples were observed. Neon plasma erosion rate and fuel retention of the micro-structured tungsten samples were almost identical to bulk tungsten samples. Keywords: Micro-structured tungsten, PFC, PFM, High heat load, Thermal cycling, Retention, Erosion, Emissivity

The Lithium Tokamak eXperiment (LTX) is a spherical tokamak device designed to study lithium plasma facing components (PFCs). The lithium coated wall of LTX has been demonstrated to produce a plasma edge with high electron temperature (200 eV or greater). Plasma density in the outer scrape-off layer (SOL) is also found to be very low, around 2 × 1017m-3, as a result of the low recycling lithium boundary. The high temperature, low collisionality region of the plasma extends into the SOL. The recent upgrade to LTX-β includes installation of a neutral beam, which will provide further heating and fueling of the core plasma. Core and edge diagnostics will also be expanded. As part of this expansion, a Retarding Field Energy Analyzer (RFEA) has been developed for the SOL of LTX-β. Measurements of the ion temperature, ion energy distribution, and the local space potential will be performed in the SOL plasma using this RFEA. Upgraded high field side (HFS) and low field side (LFS) Langmuir probes will replace existing triple probes so that higher electron temperatures can be more reliably measured. The HFS probes are also positioned to give radial and vertical gradient measurements. The design of the RFEA will be presented, along with calibration data.Since a high temperature, low collisional edge is expected for LTX-β, with a high mirror ratio near the LCFS (around 4), the majority of particles in the SOL will be mirror-trapped. Trapped particle effects will therefore become significant in the physics of the SOL plasma, and warrant further theoretical investigations. Here we present a theoretical study of the ambipolar potential formed in the collisionless SOL via differential loss of the electrons and ions, known as the Pastukhov potential in the literature. Numerical results will also be presented. Keywords: Lithium PFC, Scrape off layer, Ambipolar potential, Retarding field energy analyzer

The design of the F lowing LI quid T orus (FLIT) at Princeton Plasma Physics Laboratory (PPPL) is presented. FLIT will focus on the development of a liquid metal (LM) diagnostics and divertor system (without a plasma source) suitable for implementation in present-day fusion systems, such as NSTX-U. FLIT is intended to provide proof-of-concept for fast-flowing LM divertor designs for heat fluxes > 10 MW/m2. The toroidal test article (ID ≈ 0.56 m, OD ≈ 1.9 m, h ≈ 0.61 m) consists of 12 rectangular coils that can generate a centerline magnetic field of 1 T magnetic for greater than 10 s. Initially, 30 gallons Galinstan (Ga-In-Sn eutectic) will be recirculated within the test article using six jxB pumps to achieve flow velocities of up to 10 m/s across the fully annular radial test section. FLIT is designed to be a flexible machine that will allow experimental testing of various LM injection techniques, the study of flow instabilities, and electromagnetic control concepts to prove the feasibility of the LM divertors within fusion reactors.

This paper is focused on the analysis of surface tension contribution to the erosion features of tungsten resolidified surfaces and resulting material response to a large number of repetitive transient hydrogen plasma impacts. Experimental investigations of erosion processes on castellated tungsten surfaces in conditions relevant to uncontrolled ITER ELMs have been performed within powerful quasi-stationary plasma accelerator QSPA Kh-50. The surface energy load measured with a calorimeter was varied between melting (0.6 MJ/m2) and evaporation (1.1 MJ/m2) thresholds, the plasma pulse duration was 0.25 ms. Observations of plasma interactions with exposed W surfaces, analysis of dust particle dynamics and the droplets monitoring have been performed with a high-speed digital camera. Repetitive plasma loads above the melting threshold led to the formation of melted and resolidified surface layers. Networks both macro and intergranular cracks appeared on exposed surfaces. Cracks propagate to the bulk mainly transversely to the irradiated surface. The splashing of dust/liquid particles has been analyzed in the course of repetitive plasma pulses. It was revealed that mountains of displaced material at the edges of castellated units are the primary source of the splashed droplets due to the development of instabilities in a melted layer. The solid dust ejection dominates by cracking processes after the end of pulse and surface resolidification. Due to the continuously growing crack width (from value of sub-μm till tens μm) with the increasing number of pulses the initially uniform melt pool on the castellated units became disintegrated into a set of melt structures separated by cracks. After a large number of exposures the progressive corrugation of the surface occurred due to the capillary effects on exposed W surfaces. Results of simulation experiments for castellated targets and developed surface structures are compared with repetitive plasma exposures of flat tungsten surfaces. Keywords: Plasma-Materials Interaction, Tungsten, Surface Effects, Divertor Materials

One of the main aims of Wendelstein 7-X (W7-X), an advanced stellarator, is to investigate the quasi-steady state operation of magnetic confinement devices for nuclear fusion, for which power exhaust is an important issue. A dominant fraction of the energy leaving from the confined plasma region will be removed by 10 so-called island divertor units, which are designed to sustain a maximum heat flux of up to 10 MWm−2. An essential prerequisite for the safe operation of a steady-state device is automatic detection of hot spots and other abnormal events. Simple temperature limits in infrared (IR) thermographic images will not be enough on their own, because of plasma-generated surface coatings and other effects summarized in the following. To protect divertor elements from overheating, and to monitor power deposition onto the divertor elements, near real-time hotspot detection algorithms for the analysis of carbon plasma facing components (PFCs) were implemented and tested in the GLADIS facility.One of the difficulties in hotspot detection in a carbon-based machine is the deposition of plasma impurities as layers with a reduced thermal connection to the underlying bulk material. We have developed and successfully tested a method to classify surface layers and benchmarked the performance of the method with the Tore Supra IR data operating with actively cooled carbon PFCs. The surface layers can be detected in a steady plasma discharge during the initial rise and decay in temperature when a strike line touches parts of the divertor or wall. It can also be detected by modulating electron cyclotron resonance heating (ECRH) input power. This feature allows detection of overheated areas while reducing false positives. For the recent operational campaign, inertially cooled test divertor units (TDU) were installed to prepare for steady-state operation with water-cooled divertor units. Automatic, near real-time detection of hot spots and identification of surface layers in the W7-X divertor are presented. Results are compared with a best fit estimate of the heat transmission coefficient α which is used to calculate heat flux onto the divertor in the presence of surface layers. Keywords: W7-X, Divertor, Plasma facing components, Heat load, Infrared, Surface layers, Tore supra

Three dimensional (3D) distribution of impurity emission intensity is obtained from two dimensional (2D) image using newly developed single field-of-view tomography scheme. Spectrometer with multi-channel fiber array views the entire edge region of stochastic magnetic layer of LHD, including divertor plates, divertor legs, the stochastic layer, and the last closed flux surface. The tomography scheme introduces new regularization terms in Lagrangian function, based on the transport feature in magnetized plasma that the emission assumed to be smooth along the magnetic field due to the very fast parallel transport compared to the transport across the magnetic field. Feasibility of the scheme is checked with test emission distributions. The scheme is applied to the experimental data in LHD, where the impurity emission distribution of different charge states and of different electron densities are obtained in relation to the magnetic field structure in the edge stochastic layer.

Ammonia molecules, which will be formed in a nitrogen-seeded divertor plasma, show significant catalytic effects for volumetric recombination of hydrogen ions. This new Molecular Assisted Recombination (MAR), named the Hydronitrogen-MAR (HN-MAR), is based on the charge/ion exchange with NH3 and subsequent volumetric recombination of NH3+ and NH4+. In our PISCES-E RF plasma device, Ar is mixed with ND3/D2 gasses to realize a high electron density plasma [ne ∼ 1017 m−3] which allows us to see the importance of the dissociative recombination reactions which become comparable with wall loss reactions in our chamber. As increasing the plasma density, a calibrated Electrostatic Quadrupole Plasma analyzer measured dropping of ND4+ density fractions which can be explained reasonably by the existence of its dissociative recombination process in the plasma as a dominant reaction. Keywords: Ammonia, MAR, Detachment, Charge exchange, Recombination, Combined ion mass and energy analyzer

Previous simulation results for the tokamak ASDEX Upgrade have shown that Magnetic Perturbation (MP) fields could lead to a decrease of the peak heat flux onto the divertor targets. However, experimentally, such a decrease has not been observed. To understand this effect, a well diagnosed ASDEX Upgrade discharge was simulated with the three dimensional transport code EMC3-EIRENE, where an ad hoc model was applied to mimic the screening field created by the plasma in response to the external MP field. The simulations show that only for sufficiently weak screening a broadening of the plasma wetted area occurs. The heat flux broadening effect was then studied more systematically by simulations where heat and particle diffusion coefficients were scanned for three different MP field strengths. The scans show that only for sufficiently low diffusion coefficients and high MP field strengths, a significant decrease of the peak heat flux can be achieved. Furthermore, the effect of a kink response, which is commonly observed during ELM suppression discharges in ASDEX Upgrade, was investigated with EMC3-EIRENE simulations. The simulation results indicate, that while a kink response can deform the midplane profiles strongly, the effect on the power decay length may still be small. Keywords: ASDEX upgrade, Magnetic perturbations, Divertor heat flux, MSC: 00-01, 99-00

Tungsten is the material of choice for plasma-facing components in the divertor region of future plasma-burning tokamak fusion reactors. However, under low energy helium irradiation at elevated temperatures, significant surface morphology changes are expected, including pores, blisters, and fuzz. Dispersion-strengthened tungsten materials with small additions of transition metal carbide dispersoids have been proposed as an alternative to pure tungsten, as they have shown enhanced thermomechanical properties and possible radiation damage tolerance. However, their response to low energy helium irradiation has yet to be fully elucidated. In this work, dispersion-strengthened tungsten alloys containing 1–10 wt.% tantalum carbide, titanium carbide, or zirconium carbide were exposed to 250 eV helium ions at 600 and 800 °C to a 1 × 1020 cm−2 fluence to understand the early-stage irradiation response to helium bombardment. At 600 °C, nanostructuring is only observed on titanium carbide particles. As the temperature is raised to 800 °C, pores and ripples were developed on tungsten grains for all samples; fiber-form structures and isolated tendril growth is observed only on titanium carbide particles. Minimal surface morphology changes were observed on tantalum carbide and zirconium carbide particles. X-ray photoelectron spectroscopy of the alloyed specimens post-irradiation at 800 °C indicates the formation of zirconium and titanium oxides on the surface. Potential thermodynamic, sputtering, and composition-based formation mechanisms behind the novel nanostructuring and chemistry changes of the complex materials are discussed. Keywords: Dispersion-strengthened tungsten, Tungsten, Helium damage, Tungsten fuzz

To explore the influence of drifts on ion flows in the DIII-D divertor and Scrape-Off-Layer (SOL), a calibrated Coherence Imaging System (CIS) views the lower divertor. A new tuneable calibration laser and precision wavemeter provide the rest wavelength for CIII 465 nm plasma emissions resulting in ≤ 1 km/s uncertainty in ion velocities that are in the range of 30 km/s in the divertor plasma. The 2-D line-integrated velocity images have been tomographically inverted to obtain maps of ion velocity. Both He + and C 2 + ion velocity maps were measured in both directions of the toroidal magnetic field BT, and C 2 + images were obtained in attached and detached L- and H-mode plasmas. The measured C 2 + velocities are in the range of 20–30 km/s and are compared with UEDGE modeling with full drifts. While He II as a main ion is expected to be entrained in the plasma flow, the C 2 + impurity flow involves the competition between the frictional and ion temperature gradient forces. Both the CIS data and UEDGE modeling agree that the C 2 + flow is towards the divertor plates, but flips with the direction of BT. In general, the experimental C 2 + emission and flow profiles have a wider spatial extent than the UEDGE modeling.

ASDEX Upgrade is planning the installation of a pair of in-vessel coils in order to study alternative divertor configurations, like an X-divertor (XD) or a Snowflake (SF) in a machine with a high heating power compared to its size. Two designs for the technical realization of these coils are currently proposed, one including z-shaped joints and one with a single spiral conductor. The magnetic fields produced by these conductors intrinsically have a non-axisymmetric component. In particular in the XD configuration, where very shallow field line incidence angles occur, these 3D fields have a strong impact on the power distribution on the outer divertor plate. The 2D and 3D plasma transport is studied by the SOLPS and initial EMC3-EIRENE code simulations, respectively, in order to support the decision making processes of the engineering design. Although several aspects still require further attention, this paper gives a status report. The strong toroidal asymmetries observed in the power deposition pattern for the spiral conductor seem to be strongly reduced, when highly dissipative divertor conditions are reached. The asymmetry for the z-joint design is substantially smaller than the spiral conductor even for low-dissipative conditions. Keywords: Power exhaust, Alternative divertor, ASDEX upgrade, EMC3-EIRENE, Tokamak, X-divertor

The 2016 DIII-D DiMES tungsten (W) leading edge experiment in support of ITER studied heat loads in helium (He) plasmas with ELMs and compound ELMs (C-ELMs). The regime was close to the threshold for L-mode to H-mode transitions. In shots with ECH, C-ELMs produced in H-L back-transitions dominated the transient particle exhaust. Their duration (10–20 ms) was much longer than regular ELMs, and their heat loads higher by 2–3 times. The C-ELMS reduced the plasma density significantly, and some triggered (automated) gas puffing. This regime, with high particle and heat exhaust, may have relevance for He plasmas in ITER's start-up phase. Regular ELMs occurred during neutral beam (NB) powered shots. This paper describes new analyses on heat loads during and between C-ELMs in ECH-powered shot 166843. The results are generally consistent with a geometric solution for the parallel heat load at the plasma edge striking the leading edge. The power in the C-ELMs is sufficient to cause a temperature rise of ∼

Experiments at DIII-D use different divertor geometries to measure the effect of divertor closure on detachment onset. We compare matched H-mode discharges in the closed (upper) divertor and the open (lower) divertor and use as a detachment onset metric the rollover in peak ion flux to the target, Jsat, as a function of upstream pedestal top density, ne,ped. Measurements of electron temperature in the lower divertor confirm that this metric captures the transition from attached to detached divertor for this data set. Experiments at two different neutral beam injection powers demonstrate an ≈ 20–40% decrease in pedestal density at detachment onset in the closed case compared to the open. High-resolution edge density and temperature measurements and power balance analysis allow us to determine the difference in upstream separatrix density ne,sep at detachment onset, which we find to be ≈ 15–25% lower in the closed divertor case. The effect of closure on detachment seen in separatrix density is hence only about 60–75% the effect seen in pedestal density, with the remainder due to differences in ne,sep/ne,ped. Future devices are expected to have a scrape off layer that is opaque to neutrals, and hence will not have the same pedestal fueling to affect ne,sep/ne,ped; this suggests that the effect of closure on detachment onset seen in future devices will be less than that seen in current experiments, and closer to the magnitude of the effect we see here in the separatrix for a similar degree of divertor closure. Keywords: Tokamak, Divertor, Detachment, Fusion

Zircaloy-2 and Zircaloy-4 were neutron irradiated at 358 °C in the high flux isotope reactor (HFIR) to a neutron fluence of 29.6 × 1024 n/m2 (E > 1 MeV) and the resultant irradiation hardening was measured by tensile testing. Irradiation at the relatively high temperature of 358 °C resulted in about 25–45% less hardening than observed in the literature for irradiations at 260–326 °C. The lower irradiation hardening was shown to result from a coarser loop distribution. The measured irradiation hardening was closely related to the 〈a〉 loop size and number density (ND) using an Orowan hardening model when the loops were assumed to be strong barriers like voids. In this work Atom Probe Tomography (APT) was used to detect the clustering of Fe, Sn, and Cr at features consistent with loops. The tendency for clustering of Fe, Sn, and Cr at loops could result in the apparent high barrier strength for these features. These clusters are also an additional hardening barrier. However, local variation in the distribution of clusters is observed from region to region in the irradiated material that appears to be related to factors such as the proximity to precipitates, local variations in microstructure, and/or microchemistry. Keywords: Zircaloy, Irradiation Effects, Hardening, Clustering

One of the options currently taken into account for the realization of the first DEMO reactor is the “water-cooled lanket”. This option implies an irradiation temperature for the blanket material in the range of 280–350 °C. Therefore, in light of the under irradiation behaviour of EUROFER, namely of the DBTT shift toward high temperature due to the low irradiation temperature embrittlement, the target of the hereby reported activities is the development of much tougher alloys, to try to tolerate the embrittlement due to the low irradiation temperature. We report in this paper the work done to optimize the toughness of Eurofer 97, increasing the normalizing temperature and maintaining a small grain size using multiple normalizing treatments. We report also the mechanical behaviour of two 9Cr1WTa type alloys, produced and tested with the same aim to find alloys more resistant to embrittlement at low irradiation temperature. Keywords: EUROFER 97, RAFM steels, Microstructure, Heat treatment, Tensile, Charpy, KLST

In order to simulate hydrogen charging and discharging cycles of mechanically loaded structures full 3D Macroscopic Rate Equation (MRE) modelling is proposed based on a finite element method (FEM). The model, implemented in the 3DS Abaqus software, uses a generalized transport equation, which accounts for mechanical fields, hydrogen transport and trapping, and their evolution with time. The influence of a-priori known thermal field has also been included. To ensure the solution convergence and the numerical stability, the trapping kinetic is introduced by using an approximation of the analytical solution the McNabb and Foster equation. Comparisons with a relevant 1D MRE code and with thermal programmed desorption (TPD) experimental results are performed on a 1D configuration to validate the model. Next, the model is used to simulate the tritium diffusion and trapping in a 2D geometry of interest in the upper plug of ITER tokamak, and results of tritium inventory are compared with an equivalent 1D calculation. Keywords: Hydrogen, Kinetic trapping, Modelling, Finite elements, Abaqus, User subroutine, Macroscopic Rate Equations